ORNL-4865

National Technical Information Service
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Ccntract No. W-5405eng-26
FlSSrON PRODUCT BEHAVIOR
IN THE MOLTEN SALT REACTOR EXPERllMEMT
E. L. Compere E. G. Bohlmann
S. S. Kirslis F. F. Blankenship
W. 8. Grimes
OAK RIDGE NATIONAL bABOFL4TORY
Oak Ridge, Tennessee 37830
operated by
UNION CAR5IDE CORPORATiON
for the
ENERGY FiESEAROH AND DEVELOPMENT /UX4lNlSTRATlON
UC-76 - WlsCten Salt Reactor %eeknology

This report is one of periodic reports in which we describe the progress of the program. Other reports
issued in this series are listed below.
OKNL-2474
ORNL-2626
OWL-2684
OKWL-2723
ORYL-2799
ORSL-2800
ORNL-2973
ORNI730 14
ORNL-3 122
ORNL-32 15
ORNE-3282
ORNL-3360
OR%-34 IO
ORNL-3519
ORNL-3626
ORNL-3408
ORNL-38 12
ORNL-3872
ORWL-3936
ORNL-4037
QRNL-4119
ORNL-4 19 I
ORNL-4254
ORNL-4344
QRWL-4396
URSL-444‘3
OKSL-4548
ORKL-4622
ORNL-4676
OKNL-4728
ORNL-4782
Period Ending January 3 1. 1958
Period Ending October 3 1, 1958
Period Ending January 3 I, I959
Period Ending April 30, 1959
Period Ending July 3 1, 1059
Period Ending October 3 I ~ 1959
Periods Ending January 31 and April 30, 1960
Period Ending July 3 1, 1960
Period Ending February 28, 196 1
Period Ending August 3 1, 196 1
Period Ending February 28, 1962
Period Ending August 3 1, 1962
Period Ending January 3 I, 1963
Period Ending July 3 1 ~ 1963
Period Ending January 3 I, 1964
Period Ending July 3 1, 1964
Period Ending February 28, 1965
Period Ending August 3 1 ~ 1365
Period Ending February 28, 1966
Period Ending August 3 1, 1966
Period Ending February 28, 1967
Period Ending August 3 I, 1964
Period Ending February 29, 1968
Period Ending August 3 I, 1968
Period Ending February 28, 1969
Period Ending August 31, 1969
Period Ending February 28, 1970
Period Ending August 3 1, 1970
Period Ending February 38, 1971
Period Ending August 3 I I 197 1
Period Ending February 29, I972

CONTENTS
ABSTRACT ............................................................................. 1
FORBWAKD ............................................................................ 1
1 . "TWODUCTION ..................................................................... 2
2 . THE MOLTEN SALT REACTOR EXPERIMENT ............................................ 3
3 . MSRE CHRONOLOGY ................................................................. 10
3.1 Operation with 235U F;ueI .......................................................... 10
3.2 Operation with 3U Fuel .......................................................... 12
4 . SOME CHEMISTRY FUXDAMENTALS ................................................... 14
5 . INVENTORY ........................................................................ 16
6 . SALTSAMPLES ............... .................................................... 19
6.1 Ladle Saxnples ..................................... ........................ 19
6.2 Freeze-Valuc: Samples .............................................................. 22
6.3 Double Wall Capsuie ..........................................................
6.4 Fission Product Element Grouping . ......................................
6.5 IVoblc-Metal Behavior .................. .......................... 27
..................................................... 28
6.7 lodine .......................................................................... 28
6.8 Tellurium ............................................. ....... ......... 72
7 . SURFACE IlbH"$SITBO?I OF FISSION PRODUC 1's BY PUMP BOWL EXPOSURE .................. 37
7.1 Cable .......................................................................... 37
7.2 Capsule Surfaces .................................................................. 37
7.3 Exposure H:xpemients ............................................................. 37
7.4 Mise ........................................................................... 41
8~ GASSAMPLES ....................................................................... 48
8.1 Freeze-Valve Capsule .............................................................. 48
8.2 Validity of Cas Samples ............................................................ 48
8.3 Double-Wall Freeze Valve Capsule .................................................... 40
8.4 Ef~eectol'Wist .................................................................... 50
9 . SURVEILLANCE SPECIMENS .................................
... 60
9.1 Assemblies I- 4 ......................
............................. 60
9.1.1 Reface ........................................
.... 60
9.1.2 Relative deposit intensity ........... .................................. 63

iV
9.2 Final Assembly ................................................................... 70
9.2.1 Design ....................... ....................................... 70
9.2.2 Specimens and flonr ........................................................ 70
9.2.3 Fission recoil ............................................................. 73
9.2.4 Salt.seekingnuclides ........................................................ '73
9.2.5 Nuclides with noble-gas precursors .......................................
9.2.6 Noble metals: niobium and mulybdenutn ........................................ 74
9.2.4 Ruthenium ........................................................
9.2.8 Tellurium ................................................................ 74
9.2.9 Io$ine ...................................................................
9.2.10 Sticking facturs ...........................................................
74
74
9.3 Profile Data .................... .............................................. 45
9.3.1 Procedure ................................................................ 75
9.3.2 Results ..................................................................
9.3.3 Diffusion mechanism relationships .............................................
78
78
9.3.4 Conclusions from profile data ................................................ 84
9.4 Other Findings 011 Surveillance Specimens ..............................................
10 . E%lhMINATION OF OFF-GAS SYSTEM COMPONCKTS OR SPECIMENS
REMOVED PKIOK TO FINAL SHUTDOWN ............................................. 91
10.1 Examination of Particle Trap Removed after Kun 7 ....................................... 91
10.2 Examination of Bff-Gas Jumper kine Removed after Run 14 ............................... 92
10.2.1 Chemical analysis .... ................... ...................... 93
10.2.2 Radiochemical analysis ................. .................... ........ 93
20.3 Examination of Material Recovered from Off-Gas Line after Run 16 ........... ............ 98
10.4 Off-Gas h e Examinations after Run 18 .................... ..................... 100
10.5 Examination of Valve Assembly from Line 523 after Run 18 .................. ..... . 104
10.6 The Estimation of FIowing Aerosol Concentrations from Deposits 011 Conduit Wid1 .............. 106
10.6. I Deposition by diffusion
10.6.2 Deposition by thermoph ................ ........... ......... 107
10.6.3 Reiarionship between ob
................ 107
10.4 Discussion of Off-Gas Line Transp ................ ..................
10.7.1 Salt constituents and salt-s
..................... 109
10.7.2 Daughters of noble gases
10.7.3 Noble metals .........
................
I 1 . FJST-OPERATION EXAMINATION OF MSRE COMPONENTS .......... .................. 112
11.1 ExarninatioIl of Deposits from the Mist Shield in the MSWE Fuel Pump Bowl ................... 112
1 I . 2 Examination of Moderator Graphite from MSRE ......................................... 120
1 1.2.1 Results of visual examination .................................................. 120
11.2.2 Segmenting of graphite stringer .................... ............... ....... 120
11 2.3 Examination of surface samples by x-ray diffraction ... .......................... 121
1 1.2.4 Milling of surface graphite samples ..................... ...................... 121
11 2.5 Radiochemical and chemical analyses of MSRE graphite .... ...................... 121
11.3 Examination of Heat Exchangers arid Control Rod Thimble Surfaces ......................... 125
11.4 Metal Transfer in MSRE Salt Circuits .................................................. 126
1 I . 5 Cesium Isotope Migration in MSRE Graphite ............................................ 127
85

il.6 Noble-Metal Fission Transport Model .
11.6. I Inventory and model . . . . . . .
11.6.2 Off-gas line deposits . . . . ‘ . .
11.6.3 Surveillance specimens . . . . . . . .
11.6.4 Pump bowl samples . . . . . . . . . .
12. SUMMARY AND OVEWVIEW . . I . . . . _ .
12.1 Stable Salt-Soluble Fluorides . . ~ . . .
12.1.1 Salt samples . . . . . . . . . . . . . . .
12.1.2 Deposition . . . . . . . . . . . . . . . . .
12.1.3 Gas samples . . . . . . . . . . . . .
12.2 Noble Metals . . . . ~ . . . . . . . . ~ . .
12.2.1 Salt-borne . . . . . . . . . . . . . . . . .
12.2.2 Niobium . , . . . . . . . . .
12.2.3 Gas-borne . . ~ . . ~ . . . . . . . . .
12.3 Deposition in Graphite and IIastelioy S . .
12.4 Iodine ............................
12.5 Evaluation ........................
. .
. .
.
. .
. .
V
. ~
. .
. .
. .
. .
. .
. .
. .
. .
~
_ .
.
. .
~ .
. . . _
. . .
. . .
. .
. .
. .
I . .
. .
. .
.
. .
. .
. .
..~..~..~..... 136
~.~..n..l..... 136
. . ..~......... 136
.............I 136
..~..L..~ . . . . . 137
. . . . . . . . . . ...‘ 138
.............I 138

Figure 2.1
Figure 2.2
Figure 2.3
Figure 2.4
Figure 3.1
Figure 6.1
Figure 6.3
Figure 6.4
Figure 6.5
Figure 7.1
Figure 7.2
Figure 7.3
Figure 8.i
Figure 8.3
FIgure 9.1
Figure 3.2
Figure 9.3
Figure 9.4
Figure 9.5
Figure 9.6
Figure 9.7
Figure 9 8
Figure 9.9
Figure 9. IO
Figure 9.1 I
Figure 9.12
Figure 9.13
Figure 9.24
Figure 9.15
Figure 9.16
vi
LIST OF FIGURES
Design flow sheet of the MSKE .....................................
Pressures, volumes, arid transit times in MSRE fuel circulating loop ....................... 4
MSRE reactor vessel ~ ........................................................ 5
Flow patterns in the MSRE fuel pump ............
Chronological outlines of MSRE operations .................... 11
Sampler-enricher schematic .....................................................
Apparatus for removing MSRE salt from pulverizer-mixer to
19
polyethylene sample bottle .................................................... 20
Freeze valve capsule ....................................... ... ...... 32
Noble-metal activities of salt samples ...... .... ........................ 30
Specimen holder designed to prevent contamination by contact with transfer tube ........... 39
Sample holder for short-term deposition test ........................................
Salt droplets on a metal strip exposed in MSRE pump bowl gas space for 10 hr ....
41
Freeze valve capsulc ........................................................... 48
Double-wall sample capsule ~ . . ... ... .................... 50
Typical graphite shapes lased in a stringer of surveillaice specimens ............... . . 60
Surveillance specimen stringer ........................ . I ........... 61
Stringer containment . . .................................. 61
Stringer assembly .................................................... ..... 62
Neutron flux and temperature profiles for core surveillance assembly ..................... 63
Schcme for milling graphite samples ... ................... 64
Firial surveillance specimen assembly .......................................
Concentration profile for ICs in impregnated CGB graphite, sample P-92 ................ 75
C:onceiitration profiles for 89Sr and I4'I3a in two samples of CGR, X-13 wide face,
andP-58 .................................. .. ...
Concentration profiles for Sr arid ' O Ba in pyrolytic graphite ...
...... 75
...... 75
Coilcentration profiles for "Nb and 95Zr in CGB graphite, X-13,
doubleexposure ........................................................
Concentration profiles for O3 Ru on two faces of the X-13 graphite
. 76
specimen, CGB, double exposure ............................................... 77
Concentration profiles for O6 Ru, 4' Ce, and '"e in CGB
graphite, sample Y-9 .... ........................... 77
Concentration profiles for d'e and 44~e in two samples of cGB graphite,
X-13, wide face, and P-38 .......................................
Fission product distribution in unirnpregnated CGB (P-55) graphite specimen ir
........... 78
in MSRE cycle ending March 25, 1968 ........................................... 49
Fission product distribution in impregnated CGB 28) graphite specimen irradiated
in MSRE cycle ending March 25, 1968 ...... .................. 80

&%.&
Figure 9.19
Figure Si. I8
Figure 9.19
Figure 9.20
Figure 9.21
Figure 9.22
Figwe 9.23
Figure 9.24
F,gure 9.25
Figure 10.1
biguie 10.2
Figure 10.3
E1gure 10.4
Figure 10.5
Plgm 10.6
Figure 10.7
Figure 10.8
Figure 1 I. I
Figure 11.2
Figure 1 I .3
Figure 11.4
Figure 1 I .5
Figure 11 .G
Figure 11.9
Figure 11.8
Distribution in pyrolytic graphite specimen irradiated in MSRE for
cycle ending hlarch 25, l?G8 ...................................................
C'raniuni-235 concentration profiles in CGB and pyrolytic graphite .......................
Lithium concentration as a function of distance from the surface: specimen Y-7 . . . ~ . .
Lithium concentration as a function of distance from the surface, specimen X- 13 ............
Fluorine concentrations in graphite sarnpk U-7, exposed to molten fuel
salt in the MSRE for iijiie months ........................................ . . 89
Fluorine concentration as a function of djstaiice from thc suzface, specimen X-13 ...........
Cornparison of lithium concentrations in samples Y-7 and X-13 .........................
%lass concentration ratio, F/I& vs depth, specimen X- 13 ............. ...
Comparison of fluorine concentrations in samples Y-'7 and X-13: a smooth line having bceii
drawn through the data points .. .................... 89
MSE off-gas particle trap . ................. ............................ 91
Deposits in particle trap Uorkmesh
........................................
Deposits on jumper line flanges after run 14 ..... ............................. 94
Sections of off-gas jumper iine flexible tubing and outiet tube after run 14 ................. 95
Deposit on flexible probe ...... .................. . L . ....... 98
Dust recovered from upstream end of' jumper Line after run 14 (16,000 w) .................
Off-gas liiic specimen holder as segmented after removal, foliowing run I8
.................
Secticln of off-gas line specimen holder showing flaked deposit,
removed after mri 18 ....... ..... .................. 104
Mist shield containing sampler cage from MSKE pump bowl ............................ I 13
Interior of mist shield ....... ....... .................. 114
Sample cage arid mist shield ..................................................... 115
Deposits on sanipler cage ......... I16
Concentration profiles from the fuel side of an MSRE heat exchanger tube measured
about 1.5 years after reactor sliutdown ........................................... 126
Coilcentration of cesium isotopes in MSRE core graphite at given distances
from fuel channel surface ...... ........................................... 127
Compartment model for nobie-metal fission transport in MSRE ......................... 129
Ratio of ruthenium isotope activities for pump bowl samples . . ~ . . . . . 132
81
84
87
87
88
88
89
'?z
99
101

Table 2.1
Table 2.2
Table 3.1
Table 4 .I
Table 5.1
Table 6.1
Table 6.2
Table 6.3
Table 6.4
Table 6.5
Table 6.6
Table 6.7
Table 6.8
Table 7.1
Table 7.2
Table 7.3
Table 7.4
Table 4.5
Table 7.6
Table 8.1
Table 8.2
Table 8.3
Table 9.1
Table 9.2
Tabie 3.3
Table 9.4
Table 9.5
Table 9.6
Table 9.7
viii
LIST OF TABLES
Physical properties of the hlSRE fuel salt .......................................... 8
Average composition of MSKE fuel salt ............................................ 9
MSWE run periods and power accumulation ........................................ 12
Free energy of formation at 650°C (AG"f. kcd) ..................................... 14
Fission product data for inventory calculations ......................................
Noble-gas daughters and salt-seeking isotopes in salt samples from MSE pump bowl
17
during uranium-235 operations ................................................. 21
Noble metals in salt samples from MSKE pump bowl during uranium-235 operation . . . . ~ . 21
Data on fuel (including carrier) salt samples from MSRE pump bowl during
uranium-233 operation ....................................................... 23
Noble-gas daughters and salt-seeking isotopes in salt samples from MSRE punlp bowl
during uranium-233 operation ................................................. 24
?Jobhe metals in salt samples from MSKE pump bowl during uranium-233 operation ......... 25
Operating conditions for salt samples taken from MSRE pump bowl during
uranium-233 operation . ............................................. 26
Data for salt samples from pump bowl during uranium-235 operation .................... 31
Data for salt samples from MSRE pump bowl during uranium-233 operation . ~ . . . . ~. . 32-36
Data for graphite and metal specimens immers~d in pump bowl .........................
Deposition of fission products on graphite and metal specimcns in float-window
38
capsule immersed for various periods in MSRB pump bowl liquid salt ................... 40
Data for wire coils and cables exposed in MSKE pump bowl ............................ 44
Data for miscellaneous capsules from MSKE pump bowl ...............................
Data for salt samples for double-walled capsules immersed in salt in the MSRE pump bowl
45
duringuranium-233 operation ................................................. 46
Data for gas samples from double-walled capsules exposed to gas in the MSRE pump
bowl during uranium-233 operation ............................................. 47
Gas-borne percentage of MSRE production rate ..................................... 51
Gas samples LJ operation .................................................... 53
Data for gas samples from MSKE pump bowl during uranium-233 operation ............ 54-59
Surveillance specimen data: first arra)' removed after run '9 ............................. 66
Survey 2, removed after run 11: inserted after run 7 .................................. 67
Third surveillance specimen survey, removed after run 14 .............................. 68
Fourth surveglance specimen survey, removed after run 18 .............................
Kelative desposition intonsity of fission products on graphite surveillance specimens
69
from find core specimen array ................................................. 71
Relative deposition intensity of fission products on Hastelloy N surveillance specimens
from final core specimen array ................................................. 42
Calculated specimen activity parameters after run 14 based on diffusion
calculations and salt inventory ................................................. 84
. ~

\.&P
Table 9.8
Table 3.9
Table 9.10
Table 10.1
Table 10.2
Table 10.3
Table 10.4
Table 10.5
Table 10.6
Table 10.4
Table 10.8
Table 10.9
Table 18.10
Table 10.11
Table 11.1
Table 11.2
Table 11.3
Table 11.4
Table 11.5
Table 11.6
Table 12.2
Table 12.3
List of milled cuts from graphite for which the fission product content could be
iw
approximately accounted for by the uranium present . ~ .
Spectrographic analyses of graphite specimens after 32,000 MWhr . . . . . ~ .
. ~ .
. ~ .
. . . . . . . . . . . . . . . . . . . . .
. ~ . . ~ . . ~ . . ~ . . ~ . .
Percentage isotopic composition of molybdenum on surveillance specimens . . ....I.. . . . . 86
Analysis of dust from MSRE off-gas jumper line . . . . . . . . . . . . . . . . . . ~ . . . . . . . . . . . , . . . . . . 06
Relative quantities of elements and isotopes found in off-gas jumper line . . . . . . . . . 94
Material recovered from MSRE off-gas line after run 16. . . . . . . . . . . . . ~ .
. . . ~ .
Data on samples or segments from off-gas line specimen holder removed following run 18 . ~ .
Specimens exposed in MSRE off-gas line, runs 15 --I8 . . . . . . . . . . . . . . . . ~ .
Analysis of depusits from line 523 . ~. . I
Diffusion coefficient of particles in 5 psig of helium . . . . . . . . ~ .
. I .
. . . . . . . . . . . . I .
Therrnophoretic deposition parameters estimated for off-gas line . . . . . . . . . . . . . ~ .
Grams of salt estimated to enter off-gas system . . . . . I . . . . ~ .
. ~ . . ~ .
. ~ . . . . . .
. . . . ~ .
Estimated percentage of noble-gas nuclides entering the off-gas based on deposited daughter
activity and ratio to theoretical value for full stripping . ~ .
. ~ .
. . . . I .
. . . . . . . . . . . . ~ .
85
86
. ~ . . I . . ~ . . 100
. . 102
. ~ . . ~ . 103
..~... 105
. . . . ~ . . . . . 106
. ~ . . . . . . . IO4
Estimated fraction of noble-metal production entering off-gas system . . . . . . . . . . . . . . . . . . . ~ I
Chemical arid spectrographic analysis of deposits from mist shield in
.
. . . . . . . . . . . . . . . ~ .
. . . . . . . . . . . . . . . .
the MSRE pump bowl . . . . ...L............................................ 117
Gamma spectrographic (,Ge-diode) analysis of deposits from mist shield in the
MSWE pump bOWl . . . . . . . . . . . . . ~ . . ~ . . ~ . . ~ . . ~ . . . . . . . . . . ~ . . . . . . . . . . . ~. . . . . . . . . 118
Chemical ar~al~ises of milled samples . . . . ~ . . . . . ~ . . ~ . . . . . ~ . . ~ . . . ................ 122
Radiochemical analyses of graphite stringer samples . . . . . . . . . . . . . . . . ~ . . ~ . . . . . . . ~ . . . . . 12.1
Fission products in MSKE graphite core bar after removal in curnularive values
of ratio to inventory I .
. ~ .
. ~ .
. ~ .
Fission products on surfaces of Hastelloy N after termination of operation expressed
Ruthenium isotope activity ratios of off-gas h e deposits . . . . . ~ .
. ~ .
. ~ .
. . . . . . . . . ~ .
. . . a .
as (obsemed dis rnin-’ crn-’)/(MSRE inventor)i/total MSRE surface area) . ~ .
. ~ .
. ~ .
. ~ .
. ~ .
. . . . ~ .
. ~ .
. ~ .
. . . ~ .
. ~ .
. ~ 110
IO
. . . . e . . ~ . . . 125
. . . . . . . . .
. ~ . . . 130
Ruthenium isotope activity ratios of surveillance specimens ........... 131
Stable fluoride fission product activity as a fraction of calculated inventory
in salt samples from 33U operation . . . . . . . . . . . . . . . . . . . . ~ .
Relativi: deposition intensities for noble metals . . . . . . . . . . . . . . I .
Indicated distribution of fission products in molten-salt reactors . . . . . . I .
. ~ .
. I .
. ~ .
. ~ .
. ~ .
. ~ .
. . . . ~ .
135
. . . . . . . . . . . . . . . . . . . 137
. . . . ~ .
, ~ .
. I . . ~ .
. I39

P
FISSION PRODUCT BEHAVIOR IN THE MOLTEN SALT REACTOR EXPERIMENT
E. E. Cornpcre E. G. Bohimanri
S. S. Kirslis F. F. Blankenship
W. R. Grimes
ABSTRACT
Essentially all the fission product data for nunierous and varied samples taken during operation of
the Molten Salt Reactor Experiment or as part of the examination of specimens rernoved after
particuiar phases of operation are repurted, together with the appropriate inventory or other basis of
cuniparison. and relevant reactor parameters and conditions. Fiasion product behavior fell into distinct
chemical groups.
The noble-gas fission products Kr and Xe were indicated by the activity of their daughters to hc
removed from the fuel salt by stripping to the off-gas during bypass flow through the pump bowl, and
by diffusion into moderator graphite, in reasonable accord with theory. Daughter products appeared
to be deposited promptly on nearby surfaces including sait. for the short-lived noblegas nuclides,
most decay occurred in the fuel sdt.
The fission product elements Rb, Cs, SI, Ea, Y, Zr, and the lanthanides all form stable fluorides
which are soluhle in fuel salt. 'These were not removed from the s:ilt, and materi:il balances were
reasonably good. An acrosol salt mist produced in the pump bowl permitted a very small aniount to be
transported into the off-gas.
Iodine was indicated (with less certainty because of soniewhat deficient material balance) also TO
remain in the salt, with no evidence of volatilization or deposition on metar or graphite surfaces.
The elements Nb, Mo, Tc, Ru, .4g, Sb, and Te are not expected to form stable fluorides under the
redox conditions of reactor fuel salt. These so-cailed noblemetal elements tended to deposit
ubiquirously on system surfaces - metal, graphite, or the salt-gas interface - so that these regions
accumulated relatively high proportions while the salt proper was depleted.
Sonic holdup prior to final deposition was indicated at least for ruthenium and teIlurium and
possibly ;ill of this group of elements.
Evidence for fission product behavior during operation over a period of 26 months with 231FJ fuel
(more than 9000 effective full-power hours) was consistent with behavior during operation using 233U
fuel over a period of about 15 months (more than 5100 effective full-power hours).
This report includes essentially all the fission product
data for samples taken during operation of the Molten
Salt Reactor Expcrinnent or as part of the examination
of specimens removed after completion of particular
phases of operation, together with the appropriate
inventory or otlm basis of cornparison appropriate to
each particular datum.
It is appropriate here to acknowledge the excellent
cooperation with the operating staff of the Molten Salt
Reactor Experiment, under P. N. Haubenreich. The
work is also necessarily based on innumerable highly
radioactive samples, and we are grateful for the con-
sistently reliable chemical arid radiochemical analyses
performed by the h alj tical Chemistry Division (J. C.
White, Director), with particular gratitude due C. E.
FOREWORD
Lamb, U. Koskela, C. K. Talbot, E. 1. Wyatt, J. HI.
Moneyhun, K. R. Kickard, H. A. Paiker, and H
Wright.
The preparation of specimens in the hot cells was
conducted under the direction of E. 111. King. A. A.
Walls, It. L. Lines, S. E. Disrnuke. k~ L. Long, D. K.
Cuneo, and their co-workers, and we express our
appreciation for their cooperation and innovative assist-
ance.
We are especially grateful to the TechriicaI Publications
Department for very perceptive and thorough editorial
work.
We also wish lo acknowledge the excellent assistance
received from our co-workers L. L. Fairchild, 3. A.
Myers, and 3. E. Rutherford.

In molten-salt reactors (or any with circulating fuel),
fission occurs as the fluid fuel is passed through a core
region large enough to develop a critical mass. The
kinetic energy of the fission fragments is taken up by
the fluid, substantially as heat, with the fission frag-
ment atoms (except those in recoil range of surfaces)
remaining in the fluid, unless they subsequently are
subject to chemicaI or physical actions that transport
them from the fluid fuel. In any event, progression
down the radioactive decay sequence characteristic of
each fission chain ensues.
In molten-salt reactors, this process accumulates
many fission products in the salt until a steady state is
reached as a result of burnout, decay, or processing.
The first four periodic groups, including the rare earths.
fall in this category.
Krypton and xenon isotopes are slightly soluble gases
in the fluid fuel and may be readily stripped from the
fuel as such, though most of the rare gases undergo
decay to alkali element daughters while in the fuel and
remain there.
A third category of elements, the so-called nobIe
metals (including Nb, Mo, Tc, Ru, Ni: Pd, Ag, Sb, and
Te) appear to be less stable in salt and can deposit out
on various surfaces.
There are a number of consequences of fission
product deposition. They provide fixed sources of
decay heat and radiation. The afterheat effect will
require careful consideration in design, and the associ-
ated radiation will make rnaintenance of ielated equip-
ment more hazardous or difficult. Localization (on
graphite) in the core could increase the neutron poison
effect. There are indications that some fission products
2
1. INTRODUCTION
(e.g., tellurium) deposited on metals are associated with
deleterious grain-boundary effects.
Thus, an understanding of fission product behavior is
requisite for the development of molten-salt breeder
reactors, and the information obtainable from the
Molten Salt Reactor Experiment is a major source.
The Molten Salt Reactor Experiment in its operating
period of nearly four years provided essentially four
sources of data on fission products:
I. Samples - capsules of liquid or gas - taken from the
pump bowl periodically; also surfaces exposed there.
2. Surveillance specimens - assemblies of materials
exposed in the core. Five such assemblies were
removed after exposure to fuel fissioning over a
period of time.
3. Specimens of material recovered from various sys-
tem segments. particularly after the final shutdown.
4. Surveys of gamma radiation using remote collimated
instrumentation, during and after shutdown. As this
is the subject of a separate report, we will not deal
with this directly.
Because of the continuing generation by fission and
decay through time, the fission product population is
constantly changing. We will normally refer all measure-
ments back to the time at which the sample was
removed during fuel circulation. In the case of speci-
mens removed after the fuel was drained, the activities
will nornzally refer to the time of shutdown of the
reactor. Calculated inventories will refer in each case
also to the appropriate time.
j.. .... j.

We will briefly describe here some of the character-
istics of the Molten Salt Reactor Experiment that might
be related to fission product behavior.
The fuel circuit of the MSRE'-4 is indicated in Figs.
2.1 and 2.2. It consisted essentially of a reactor vessel, a
circulating pump. and the shell side of the primary heat
exchanger, connected by appropriate piping. all con-
structed of Hastelloy N.5 '6Hastekloy N is a nickel-based
alloy containing about 17% molybdenum, 7% chrom-
ium, and 5% iron, developed for superior resistance to
corrosion by moiten fluorides.
The main circulating "loop" (Fig. 2.2) contained
69.13 ft3 of fuel, with approximately 2.9 ft3 more in
the 4.8-ft3 pump bowl, which served as a surge volume.
The total fuel-salt charge to the system amounted to
about 78.8 ft3 : the extra voiurne, aniounting to about
9% of the system total, was contained in the drain tanks
and mixed with the salt from the main loop each time
the fuel salt was drained from the cure.
Of the salt in the main loop, about 23.52 ft3 was in
fuel channels cut in vertical graphite bars which filled
the reactor vessel core, 33.65 ft3 was in the reactor
vessel outer annulus and the upper and lower plenums,
I 1
3
2. THE MOLTEN SALT REACTOR EXPERIMENT
6.12 ft3 was in the heat exchanger (shell side), and the
remaining 6.14 ft3 was in the pump and piping.
About 5% (65 gpm) of the pump output was recir-
culated through the pump bowl. The remaining 1200
gpin (2.67 cfss) flowed through the shell si& of the heat
exchanger and thence to the reactor vessel (Fig. 2.3).
The flow was distributed around the upper part of an
annulus separated from the core region by a metal wall
and flowed into a lower plenum, from which the entire
system could be drained. The lower plenum was
provided with flow vanes and the support structure for
a two-layer grid of 1-in. graphite bars spaced 1 in. apart,
covering the entire bottom cross section except for a
central (10 X 10 in.) area. One-inch cylindrical ends of
the two-inch-square graphite moderator bars extended
into alternate spacings of the grid. Above the grid the
core was entirely filled with vertical graphite moderator
bars. 64 in. tall, with matching round end half channels,
0.2 in.deep arid 1.2 in. wide, cut into each face.There
were 1108 full channels, and partial channels equivalent
to 32 more. Four bar spaces at the corners of the
central bar were approximately circular, 2.6 in. in diam-
eter; three of these contained 2-in. control rod thirn-

POSIT ON
10
1
2
3
4
5
c;
7
e
ij
13
LOOP CATA A1 12(:C°F, 5 psq, !2OOgpv
CIFF TPfit>!SI-
VOLUME TllllE
:33: Isecj
! l 3 4'
0 ?E 0.26
6 12 2.29
2.18 0 8i
9 T2 3.63
12 24 4.53
23 52 6 79
11 39 4.26
13? c 5'
0 l.5 0.27
4
PRESSURE
: PSIC)
23 4
13 3
69 7
44 4
43 4
43 3
35 5
25.8
_...
2c.4
ORN! -CWC 70 5152
Fig. 2.2. PP~SSU~~S, volumes, and transit times in MSRE fuel circulating hap.
ble tubes, and the fourth contained 2 removable tubular
surveillance specimen array.
At reactor temperatures the expansion of the reactor
vessel enlarged thc annulus between the core graphite
and the inner wall to about I/' in.
Model studies,' ,8 indicated that althou-& the Reyn-
olds number for flow in the tioncential graphite fuel
channels was 1000, the square-root dependence of flow
on salt head loss implied that tuibulent entrance
conditions persisted well up into the channel.
Fuel salt leaving the core passed through the upper
plenum and the reactor outlet norLIe, to which the
reactor access port was attached. Surveillance speci-
mens. the postmortem segments of control rod thimble,
and a core graphite bar were withdrawn through the
access port.
The fuel outlet line extended from the reactor outlet
nozzle to the pump entry nozzle.
The centrifugal sump-type pump operated with a
vertical shaft and an overhung impeller normally at a
speed of 1160 rpm to deliver 1200 gpm to the discharge
line at a Iiead of 49 ft, in addition to internal
circulation in the pump bowl, described below, amount-
ing to 65 gpm. Because many gas and liquid samples
were taken from the pump bowl, we \di outline here
some of the relevant structures and flows. These have
been discussed in gieater length by Engel, Haubenreich,
and Hout~eel.~

e... . . .
GRAPH TE SAMPLE ACCESS POF?T
LCIRGF GRAPHITE ;AMPIES
C@RF CE\TF?ING $HIE
GRCPHiTE - U03EQ8
STRlhGEE -
REOCTOE VESSEL A
l
BNTI-SWIRL VbNES
.-
LlNt -
5
Fig. 2.3. MSRE reactor vessel.
F-t *I61 E
CON1 R@L
7 . OLrLFT STRAINER
SUPPORT GRID
-COHt WILL COOLING PVUUI

Some of the major functions of the pump and pump
bowl were:
1. fuel circulation pump,
2.
a.
4.
5. fission gas stripper,
6. gas addition point (helium, argon, oil vapor),
7.
8.
9. salt sample point,
10. gas sample point,
11.
12.
liquid expansion or surge tank,
point for removd and return of system overflow,
system pressurizer,
holdup and outlet for off-gas and purge gas,
fuel enricher and chemical addition point,
point for contacting specimen surfaces with liquid
or gas during operation.
point for postmortem excision of some system
surfaces.
The major flow patterns are shown in Fig. 2.4.
Usually the pump bowl, which had a fluid capacity of
4.8 ft3, was operated about 60% full. Although the
overficw pipe inlet was well above the liquid level and
CRPSUCE
CRCE
OVERFLOW
PIP€
UFFGAS
LINE
6
was protected from spray, overflow rates of several
pounds per hour (0.1 to 10) resulted in the accumulation,
in a toroidal overflow tank below the pump, of
overflow salt, which was blown back to the punip bowl
at the necessary intervals (hours to weeks). The
overflow tank was connected to the main off-gas line,
but because the pump bowl overflow line extended to
the bottom of the overflow tank, little or no off-gas
took this path except when the normal off-gas exit
from the pump bowl had been appreciably restricted.
It was desirable’ to remove as much of the xenon and
krypton fission gases as possible, particularly to mitigate
the hi$ neutron poison effect of 13’xe.
Consequently, about 50 gpm of pump discharge liquid
was passed into a segmented toroidal spray ring near the
tog of the pump bowl. Many and Y8-in. perforations
sent strong jets angled downward spurting into
the liquid a few inches away, releasing bubbles,
entraining much pump bowl gas - the larger bubbles
of which returned rapidly to the surface - and
vigorously mixing the adjacent pump bowl gas and
liquid. An additional “fountain” flow of about Z 5 gpm
came up between the volute casing seal and the inapeller
shaft. Other minor leakages from the volute to the
pump bowl also existed. At a net flow of 65 g p (8.7
SUCTION
Fig. 2.4. Flow patterns in the MSRE fuel pump.
SHAFT
ORNL-DWG 69- 101 7214
CLEAN GAS
\ .>

cfm) into a pump bowl salt volume of 2.9 ft3, the
average residence time of salt in the pump bowl was
about 20 sec.
The pump bowl liquid flowed past skirts on the
volute at an average velocity of 0.1 1 fps, accelerating to
1.9 fps as the salt approached the openings to the pump
suction. Entrained bubble size can be judged by noting
that bubbles 8.04 cm (0.016 in.) in diameter are
estimated by Stokes's law to rise at a velocity of 0.1 fps
in pump bowl salt.
The salt turbulence also provided an underflow entry
to the spiral metal baffle surrounding the sample
capsule cage. The baffle, curved into a spirai about 3 in.
in diameter, with about one-fourth of a turn overlap,
extended from the sample transfer tube at the top of
the pump bowl, downward through both gas and liquid
phases, to the sloping bottom of the pump bowl, with a
half-circle notch on the bottom near the pump bowl
wall to facilitate liquid entry. Subsequent upflow
permitted release of associated gas bubbles to the vapor
space, with liquid outflow through the '/8-iri. spiral gap.
Around the entry from the sample transfer tube at
the top of the pump bowl was a cage of five vertical
'/4-in. rods ternlinating in a ring near the bottom of the
pump bowl. Sample capsules, specimen exposure devices,
or capsules of materials to be dissolved in the fuel
were lowered by a steel cable into this cage for varying
periods of time and then withdrawn upward into the
sample transfer tube to be removed. Normally (when
not in use) a slight gas flow passed down the tube, clue
to leakage of protective pressurization around closed
block valves (gas was also passed down the transfer tube
during exposure of many of the above-mentioned
items).
Gas could enter the sample baffle region f~orn the
liquid and by diffusion via the spiral gap. The rate of
passage has not been determined, but some evidence
will be considered in connection with gas saniples.
Purge gas, normally purified helium, entered the
pump bowl gas space through the annulus between the
rotating impeller shaft and the shield plug, normally at
a rate of 2.4 std literslmin. Some sealing oil vapor, of
the order of a few grams per day, is indicated to have
entered by this path. Two bubbler tubes (0.39 std
Iiter/min each) and a bubbler reference line (0.15 std
liter/min) also introduced gas into the pump bowl. With
m average pump bowl gas volume of 1.9 ft' at 5 psig
and 65O"c, a flow of 3.3 std Iiterslmin corresponds to a
gas holdup time of about 6.5 min.
In order to prevent spray from entering the overflow
line or the two 'I2 -in. off-gas exit lines in the top of the
pump bowl, a sheet metal skirt or roof extended across
the pump bowl gas space from the central shaft housing
to the top of the toroidal spray ring. That same aerosol
salt or organic mist stili was borne out of the pump
bowl was indicated by the occasional plugging of the
off-gas line and by the examination of materials
recovered from this region, to be discussed in a
subsequent section.
The areas of the Mastdloy N surfaces exposed to
circulating salt in the hlSRE fuel loop were given' ' as
follows:
Pump 30 ft2
Piping 45 ft2
Heat exchanger 346 ft'
Reactor vessel 431 ft'
852 ft2 (0.7315 x lo6 cm2)
The areas of the graphite surfaces in the core of the
MSRE are estimated from design data' to be:
Fuel channels 132.35 m2
Tops and buttoms 3.42 m2
Contact edges 80.25 ma
Support lattice bars 8.95 m2
224.94 m2 (2.2497 X IO6 cm2)
Thus the total surface area of the MSRE fuel loop Is
3.041 x cm2.
Properties of the MSRE fuel salt have been given by
Cantor,' ' Grimes.' ' and Thoma.' Some of these are
given in Table 2.1. As given by Thoma,' the average
composition of the fuel (as determined by chemical
analysis) was that shown in Table 2.2.
The power released in the reactor as a result of
nuclear fission was evaluated both from heat balance
clata' 9' and from changes in isotopic camposition.'
An originally assigned full power of 8 EIIW, corrected
for various small deviations in fluid properties and
instrument calibrations, gave a new heat balance fuU
power of 7.65, and the value based on isotopic changes
was 7.4 MW. Uncertainties in physical and nuclear
properties of the salt and in reactor instrument calibration
are sufficient to account for the difference.
At a reactor power of 7.4 MW the mean power
density in the circulating fuel was 3.6 &V per cubic
centimeter of salt. About 88% of thelfissions occurred
in the core fuel channels, about 6% in the upper head,
and 3% each in the lower head and in the outer
downflow annulus.' The average thermal-neutron flux
in the circulationg fuel was about 2.9 X 10' ' neutrons
sec-' for 2 3 5 fuel ~ at full power. The thermal-neutron
flux was about 3 X 18l3 neutrons cmP2
sec-' near the center of the core and declined both

Viscosity
Table 2.1. Physical properties of the MSRE fuel salt
Property Value Estimated precision
Thermal conductivity
Electrical conductavnty
Liquidus temperature
Keat capacity
Liquid
Solid
Density
Liquid
Ex pansivit y
Compressibility
Vapor pressure
Surface tension
Solubility of He, Kr, Xe
Isuchoric heat capacity, Cv
o(centipoises) = 0.116 exp [ 3755/T("K)]
0.010 watt cm-I "6-8
K = -2.22 f 6.81 X BO-3T("C)a
434°C
Sonic velocity
500"&': pz 3420 KXI/SeC
600°C: y = 33 10 m/sec
700°C: p =: 3200 m/ser
Thermal diffusivity
500°C: D = 2.0g x 10-3 im'isec
600°C: D = 2.14 X crn2/sec
700"~: D = 2.18 x cma/sec
Kinematic viscosity
500"~: Y= 7.4, x cm2/sec
600°C: I.'> 4.36 X IO-' cin'lsec
700"~: Y= 2.8, x IO-.' cni'isec
PraIidtl RUInbeK
500°C: PT = 35.6
600°C: PI = 20.4
900°C: Pr = 13.E
500 6.6 0.13 0.03
600 10.6 0.55 0.17
700 15.1 1.7 0.47
ROO 20.1 4.4 2.0
x 10 -R moles cm-' melt atm-'
500 0.489 16.2 6.91 1.17
600 0.482 15.9 6.81 1.18
780 0.475 15.7 6.72 1.20
*3%
-93%
fl%
f18%
Factor 3
Factor 50 from 500 to 700°C
+30, -10910
2 Factoa 10
'Applicable over the temperature range 530 to 45CF'"C. The va!ue of electrical conductivity given hcre was estimated by
G. D. Robbins and is based on the assumption that ZrF4 arid UF4 behave identically with ThF4; see 4;. D. Robbins and
A. S. GaHanter,MSR kog~am Semiennu. hog. Rep. k~g. 31. 1970, ORNL-4548, p. 159; ibid., ORNl.4622, p. 10B.

9
Table 2.2. Average composition of MSRE fuel dt
Runs 4- 14‘ Runs 16-2Qb
LIF, mole 75 64.1 + 1.1 64.5 2 1.5
&E’?, mole % 30.0 ? 1.0 30.4 ’ 1.5
ZrE’4. mole 76 5.0 t 0.19 4.90 t 0.16
UY.4 , mole 5%
Cr, ppm
0.809 f 0.024 0.134 2 0.004
64 ? 13 (range 35-80) 80 2 14 (range 35-100)
Fe, win 130 ? 45 157 + 43
Ni, ppm 64 f 61 46 2 14
‘Operation with 2 3 5 fuel. ~
’Operation with 233U fuel.
radiafly and axially to values about 18% of this near the
graphite periphery. The fast flux was about three times
the thermal flux in most cure regions.
B. E. Prince’
computed the central core flux for
233U to be about 0.8 X BO1 neutrons cm-’ sec-’per
megawatt of reactor power. or about 6 X IO’ at full
power. The relatively higher flux fur the a33U fuel
resuits from the absence of ’ 38U as well as the greater
neutron productivity of the ’ ’ U.
U fuelq ’ ’Pu
Across the period of operation with
was formed more rapidly than it was burned, and the
concentration rose until about 5% of the fissions were
coritributed by this nuclide. During the 3U operations,
the plutonium concentration fell moderately but
was replenished by fuel addition. The resultant effects
on fission yields will be discussed later.
1. P. N. Haubenreich and J. R. Engel, “Experience
with the Molten Salt Reactor Experiment,” Nucl. Appl.
Teciinol, 8, 118-37 (February 1978).
2. R. e. Robertson, MSRE Design and Operatioris
Report, Part 1. Description of Reactor Design, ORNL-
TM-728 (January 1965).
3. J. R. Engel, P. N Haubenreich, and A. Houtzeel,
Spray, Mist, Bubbles, and Foam in the Molten Salt
Reactor Experiment, ORNLTM-3027 (June 1940).
4. W. B. McDonald, “MSRE Design and Construction,”
MSR Frogram Semiannu. Brog. Rep. July
31,1964, ORNk-3408, pp. 22-83.
5. H. E. McCoy et al., “New Developments in
Materials for Molten-Salt Reactors,” IVMCI. Appi. Tech-
nol. 8, 156-69 (February 1WQ).
~-___
6. A. Taboada. “Metallurgical Developments,” MSR
Program Semiannu. I9rog-r. Repp. JuZy 31, 1964, OWL-
3708, pp. 330-72.
’7. D. Scott, Jr., “Component Development in Support
of the MSRE,” MSR Frograni Senz2nnu. Prop.
Rep. Jui‘y 31, 1964, ORNL-3788, pp. 167- 90.
8. R. J. Kedl. Fluid Qvnckmic Studies of the Molten
Sult Reactor Experiment [MSRE) Core, OWL-
ThI-3229 QNov. 19, 19’30).
9. J. R. Engel and R. 63. Steffy, Xeum Behavior in
the Molten Salt Reactor Eq?erimeizt, ORX-TM-3464
(October 1971).
10. J. A. Watts and .I. R. Engel, “Hastelloy N Surface
Areas in MSFE,” internal memorandum MSR-69-32 to
R. E. Thoma, kpr. 16, 1969- (Internal document - no
further dissimination au thhorized.)
1 1. S. Cantor, Physical Properties of Molten-Salt
Reactor Fi4eE, Cool~tit and Flus11 Salts, ORNL-TM-23 I6
(August 1968).
12. W. R. Grimes, “Molten Salt React01 Chemistry,”
Nw~. AP@. Tcdmtal. 8(2), 139- 55 (February 1970).
13. R. E. Tlionia, Chemical Aspects of MSM Operatiun,
ORNL-4658 (December 1971).
14. 61. H. Gabbard, Reactor Power Measurement and
Heat Transfer Performance in the MSlE, OWL-Th.l-
3082 (May 1971).
15. C. H. Gabbard and P. N. Haubenreich, “Test of
Coolant Salt FIowmeter and Conclusions,” MSR Program
Semiannu. Progr, Rep. Feb. 28, 6971, OHPNE-
4676, pp. 17-18.
16. J. R. Engel, “NucEear Characteristics of the
MSRE,” MSR Program Semdariw. Bogr. Rep. July 31,
1965, OWL-3708, pp. 83-1 14.
17. €3. E. Prince, “‘Other Neutronic Characteristics of
ruISRE with ’ u ~uel,” MLYR fiogarn Semiannu.
Progr. Rep. Aug 31, 1967, OW&-4191, pp. 54-61.

A sketchy chronalugy of the MSRE, with an eye
toward factors affecting fission product measurements,
will be given below. More complete details are available.
3.1 Operation with u Fuel
The MSRE was first loaded with flush salt on
November 28, 1964; after draining the flush salt, 452
kg of carrier salt ('LiF-BeF? -ZrF4 ~ 62.4-32.3-5.3 mole
5%; mot. wt 40.2) was added to a drain tank fullowed by
235 kg of 7LiF-23sUF4 eutectic salt (72.3-27.7 mole
%, mol. wt 105.7) in late April. Circulation of this salt
was followed by addition of 7kiF-235UF4 (93% en-
riched) eutectic salt beginning on May 24, 1965.
Criticality was achieved on June 1: 1965. Addition of
enriched capsules of ' LilF-' UE4 eutectic salt con-
tinued throughout zero-power experiments, which in-
cluded controlled calibration. The loop charge at the
beginning of power operation consisted of a total of
4498 kg of salt (nominal composition by weighi, 'Li,
11.08%; We. 0.35%; Zr, 11.04%; and U, 4.6285%). with
390 kg in the drain tank (ref. 2, Table 2.15).
Operation of the MSRE was commonly divided into
runs, during which salt was circulating in the fuel loop;
between runs the salt was returned to the drain tanks,
mixing with the residual salt there.
Run 4, in which significant power was first achieved,
began circulation in late December 1965; the approach
to power has been taken arbitrarily as beginning at
noon January 23> 1966, for purposes of accounting for
fission product production and decay.
Significant events during the subsequent operation of
the MSRE until the termination of operation on
December 12, 1969, are shown in Fig. 3.1. The time
period and accumulated power for the various runs are
shown in Table 3. I.
Soon after significant power levels were reached,
difficulty in maintaining off-gas flow developed. De-
posits of varnish-like material had plugged smdi pas-
sages and a small filter in the off-gas system. A small
amount of oil in the off-gas hoidup pipe and from the
pump had evidently been vaporized and polymerized by
the heat and radiation from gas-borne fission products.
The problem was relieved by instalhtion of a larger and
more efficient filter downstream from the holdup pipe.
On resumption of operation in April 1966, full power
was reached in run 6 after a brief shutdown to repair an
electrical short in the fuel sampIer-enricher drive. The
first radiochemical analyses of salt samples were re-
ported for this run. Run 7, which was substantially at
full power, was terminated in late July by failure of the
blades and huh of the main blower in the heat removal
system. While a replacement was redesigned, procured,
and installed, the array of surveillance specimens was
removed, and examinations (reported later) were made.
Some buckling and cracking of the assembly had
occurred4 because movement resulting from differenrid
expansion had been inhibited bp entrapment and
freezing of salt within tongue-and-groove joints. Modifications
in the new assembly permitted its continued
use, with removals after runs 11 14, and 18, when it
was replaced by an asseanbiy of another design.
Run 8 was halted to permit installation of a blower;
run 9, to remove from the off-gas jumper flange above
the pump bowl some flush salt deposited by an overfill.
During run 3 an analysis for the oxidation state of the
fuel resulted in a U3+/U4+ value of~.l%. Because values
nearer 1% were desired, additions of metallic beryllium
as rod (or powder) were made' using the samplerenricher,
interspersed with some samples from time to
time to determine U3'/U4".
During run 10 the first "freeze-valve" gas sample was
taken from the pump bowI. The series of samples b e ~ n
at this time will be discussed in a later section. Run IO
operated at full power for a month, with a scheduled
termination to permit inspection of the new blower.
Run 11 lasted for 102 days, essentially at full power,
and was terminated on 'schedule to permit routine
exarninaticns and return of the core surveillance specimen
assembly. During this run a total of761 g of' 35U
was added (as EiP;-UF4 eutectic salt) without difficulty
through the sampler-enricher, while thc: reactor was
in operation at fulI power. After completion of maintenance
the reactor was operated at full power during
run 12 for 42 days. During this period, 1527 g of
' 35 eT was added using the sampler-enricher. Beryllium
additions were followed by samples showing
1J3'/U4' of 1.3 and 1.0%. Attempts to untangle the
sampler drive cable severed it, dropping the sample
capsule attached to it, thus terminating run 12. The
cable latch was soon recovered: the capsule was subsequently
found in the pump bowl during the final
postmortem examination. Run lib commenced on
September 20) after some coolant pump repairs, and
continued without fuel drain for 188 days; the reactor
was operated subcritical for several days in November
to permit electrical repairs to the sampler-enricher.
Reactur power and temperature were varied to determine
the effect of operating conditions on ' 35~e stripping.' During run 14 the first subsurface salt
samples were taken using a freeze-valve capsule.

SALT IN
FUEL LOPP
(966
1467
J
J
A
S
J
J
A
POWER
0 2 4 6 8 K J
FUEL POWER (Mw)
FLUSH 0
DYNAMICS TESTS
NVESTIGATE
OFFGAS PLUGGING
REPLACE VALVES
AND FILTERS
RAISE POWER
REPAIR SAMPLER
ATTAIN FULL POWER
CHECK CONTAINMENT
FULL- POWER RUN
- MAIN BLOWER FAILURE
REPLACE MAIN BLOWER
MELT SALT FROM GAS LINES
REPLACE CORE SAMPLES
TEST CONTAINMENT
RUN WITH ONE BLOWER
> INSTALL SECOND BLOWER
} ROD CUT OFFGAS LINE
CHECK CONTAINMENT
30-doy RUN
AT FULL POWER
} REPLACE AIR LINE
DISCONNECTS
SUSTAINED WERATION
AT HIGH POWER
RPLACE CORE SAMPLES
TEST CONTAINMENT
REPAIR SAMPLER
11
SALT IN
FUEL LOOP POWER
0
N
0 -P I
FUEL
FLUSH 0
1
0 2 4 6 8 4 0
POWER (Mw)
Fig. 3.1. Chronological outline of MSRE operations.
ORNL-LWG 69-7293R2
XENON STRIPPING
EXPERIMENTS
t REPLACE
INSPECTION AN0
MAINTENANCE
CORE SAMPLES
TEST AND MOOIFY
FLVORINE MSPOSAL
SYSTEM
PROCESS FLUSH SALT
PROCESS FUEL SALT
LOAD URANIUM-233
REMOVE LOADING MVICE
233U ZERO-POWER
PHYSICS EXPERIMENTS
INVESTIGATE FUEL
SALT BEHAVIOR
CLEAR OFFGAS LINES
REPAIR SWPLER WO
CONTROL ROD DRIVE
233U DYNAMKS TESTS
INVESTIGATE GAS
IN FUEL LOOP
HIGH-POWER OPERATION
TO MEASURE 233U pc/wa
INVESTIGATE COVER Gps.
XENON, AND FISSION
FRODUCT BEHAVIOR
ADD PLUTONIUM
IRRADIATE ENCAPSULATED U
MAP F.P. DEPOSITON WITH
GAMMA SPECTROMETER
MEASURE TRITIUM,
SAMPLE FUEL
REMOVE CORE ARRAY
PUT REACTOR IN STANDBY

Run Date started Date drained Run hours
Cumulative total Cumulative effective full-
hoursa power hours
4 1-23-6Ga 1-2666 80P 80.8 4
5 2-13-66 2-1666 55 .O 564.5 5
6A 4-8-66 4-22-66 342.1 2,146.3 54
6B 4-25-66 4-23-66 107.5 2,312.7 115
6C 5-8-66 5 -2 8-6 6 475.0 3,005.0 374
7A 6-12-56 6-28-66 348.1 3,711.4 6 84
75 6-30-66 7-23-66 553.0 4,355.7 1,055
Surveillance specimen assembly removed. New assembly installed.
8 10-8-66 10-31-66 546.1 6,747.6 1,386
9 11-7-66 1 I-20-66 301.2 4,213.0 1,545
10 12-14-66 1-18-67 824.2 8,628.5 2,262
I1 1-28-67 5-1 1-67 246 1.4 11,340.6 4,510
Surveillance specimen assembly removed. Reinstalled.
12 6-1 8-67 8-1 1-67 1217.8 13.548.3 5.566
13 9-1567 8-18-64 77.8 14,44 1.7 5,626
14 9-20-67 3-25-68 4468.2 18,997.0 9,005
Sarveillance specimen assembly removed, reinstalled. Off-gas specimen instailed.
235U removed from carrier salt by fluorination. 233U fuel added.
15 10-2-68 11-28-68 i372.1 24,956.1 9,006.5
14 12-1 2-68 12-17-68 111.0 25,404.0 9,006.5
17 1-13-69 4-1069 2085.8 28,146.1 10,489
18A 4-12-69 4-15-69 74.7 28.269.3 10,553
18B 4-16-69 6-1-69 1104.4 29,402.6 11,547
Surveillance specimen assembly removed. New assembly installed.
19 8-17-69 11-2-69 1856.7 33,098.7 12,790
20 11-25-69 12-12-69 396.7 34,055.3 13,172
I:inal drain. Surveillance specimen assembly removed. System to standby.
Postmortem, Januxry 1941. Segments from core graphite, rod thimble, heat exchanger, pump bowlril, freeze valve. System to
standb p .
“From beginning of appruach to power, taken as noon, Jan. 23, 1966. Prior circulation in run 4 not included.
After the scheduled termination of run 14. the core
surveillance specimen assembly was removed for exami-
nation and returned. The off-gas jumper line was
replaced; the examination of the removed line is
reported below. A specimen assembly was inserted in
the off-gas line.
AI1 major objectives of the * ’ U operation had been
achieved, culminated by the sustained final run of over
six months at full power, with no indications of any
operating instability, fuel instability, significant corro-
sion, or other evident threats to the stability or ability
to sustain operation indefinitely.
3.2 Operation with 3 3 ~uel ~
It remained to change the fuel and to operate with
fuel, which will be the normal fuel for a
2 3 s ~
molten-salt breeder reactor. This was accomplished
across the sunimer of 1968. The fuel, in the drain tanks,
was treated with fluorine gas, and the volatilized UF6
was caught in traps of granular NaF. Essentially all 218
kg of uranium was recovered,” and no fission pruducts
(except ”i\;b), inbred plutonium, or other substances
were removed in this way. The carrier salt was then
reduced by hydrogen sparging and metallic zirconium
treatment, filtered to remove reduced corrosion prod-
ucts, and returned to the reactor. A mixture of * s UF4
and 7LiF was added to the drain tanks, and some
UF4 was included to facilitate desired isotope ratio
determinations.
Addition of capsules using the sampler-enricher per-
mitted criticality to be achieved, and on October 8,
1968, the US. Atomic Energy Commission Chairman,

Glenn Seaborg, a discoverer of 2 3 3 ~ first , took the
reactor to significant power using U fuei.
The uranium concentration with 233kj fuel (83%
enriched) was about 0.3 mole 70. The fuel also
contained about 540 g of 239Pu, which had been
formed during the 245LJ operation when the fuel
contained appreciable U.
During the final months of 1948, zero-power physics
experiments were accompanied by an increase in the
entrained gas in the fuel. Beryllium was added to halt a
rise in the chromium content of the fue1. Some finely
divided iron was recovered from the pump bowl using
sample capsules containing magnets. During a subsequent
shutdown to combine all fuei-containing salt in
the drain tank for base-line isotopic analysis, a stricture
in the off-gas line was removed, with some of the
material involved being recovered on a filter.
At the beginning of run 17 in January 1969, the
power level was regularly increased, with good nuclear
stability being attained at full power. Transients attributed
to behavior of entrained gas were studied by
varying pump speed and other variables; argon was used
as cover gas for a time. Freeze-valve gas samples and salt
samples were taken, and a new double-wail-type sample
capsule was employed. Further samples were taken for
isotopic analysis. The lower concentrations of uranium
in the fuel led to unsuccessful efforts to determine the
U3+/U4+ ratio. However, beryllium additions were continued
as @+ concentration increases indicated.
In May 1969, rcstrictions in the off-gas lines appeared
and subsequently also in the off-gas line from the
overflow tank. Operation continued, and run 18 was
terminated as scheduled on June 1 ..
Surveillance specimens were removed, and an assembly
of different design was installed. This assembly
contained specially encapsulated uranium, as well as
material specimens. A preliminary survey of the distribution
of fission products was conducted, using a
collimated Ge(Li) diode gamma spectrometer.6 This
was repeated more extensively after run 19.
After completing scheduled routine maintenance, the
reactor was returned to power in August 1969 for run
19. Plutonium fluoride was added, using the samplerenricher,
as a first step in evaluating the possibility of
using this material as a significant component of
molten-salt reactor fuel.
At the end of run 19, the reactor was drained without
flushing to facilitate an extensive gamma spectrometer
survey of the location of fission products.
The fate of tritium in the system was of considerable
interest, and a variety of experiments were conducted
and samples taken to account fur the behavior of this
product of reactor operation.’ ,8
13
Because salt aerosol appeared to accompany the gas
taken into gas sample capsules, a few double-walled
sample capsules equipped with sintered metal filters
over the entrance nozzles were used in run 28.
After final draining of the reactor on December 13.
1969, the surveillance specimen assembly was removed
for examination, and the reactor was put in standby.
In January 1971 the reactor cell was opened. and
several segments of reactor components were excised
for examination. These included the sampler-enricher
from the pump bowl, segments of a control rod thimble
and a central graphite bar from the core. segments of
heat exchanger tubes and shell, and a drain line freeLe
valve in which a small stress crack appeared during final
drain operations. The openings in the reactor were
scaled, and the reactor crypt was closed.
References
1. P. N. Maubenreich and J. R. Engel, “Experience
with the Molten Salt Reactor Experiment,” Nucl. Appl.
Technol. 8(2), 18 -36 (1969).
2. W. E. Thonia, Chemical Aspects of MSRB Operation,
ORNL-4658 (December IVI).
3. MSR Progrum Semiunnu. Progr. Rep. (a) Jdy 31,
1964, ORNL-3708; (b) Feb. 28, 1965, 0RNL-3812:(~)
Aug. 31, 1965, QRNL-3842; (d) Feb. 28, 1966,
ORNL-3936; (e) Aug. 31, 1966, ORNL-4037; (f) Feb.
28,1967,ORNL-4119;(g)A~g. Si, 1967, ORNL-4191:
(h) Feb. 29. IY68, OKNL-4254; (i) Rug. 31. 1968,
ORNL-4344: 0) Feb. 28, 196Y, OKNL-4396; (k) Au~.
SI, 1969, ORNL-4449; (l) Feb. 28, 1990, QRNL-4548;
(.e) Rug. 31, 1970, ORNL-4622; (rz) Feb. 28, i97I1,
ORNL-4676; (0) Aug. 31, 1971, ORNL-4728.
4. W. H. Cook, “MSRE Matcrials Surveillance Testing,”
MSK Progrum Semiunizu. Bog. Rep. Aug. 31,
1966, ORNL-4037, pp. 97 -103.
5. 9. R. Engel and R. 6. Steffy, Xenon Behavior in
the Molten Salt Reactor Experiment, ORNL-TM-3464
(October 1971).
4. A. Houtzeel and F. F. Dyer, A Study ofB;iSsion
?roducts in the Molten Sdt Reactor Experimerrt by
Gamma Spectrometry, ORNL-TM-3 15 1 (August 1972).
7. P. N. Haubenreich, (a) “Tritium in the MSE:
Calculated Production Rates and Observed Amounts,”
QRNL-CF-70-2-7 (Feb. 4, 1970); (6) “A Review of
Production and Observed Distributions of Tritium in
MSRE in the Light of Recent Findings,” OWL-
CF-41-8-34 (Aug. 23. 1971). (Internal d~ci~ments - nu
further dissemination authorize d).
8. R. B. Briggs, “Tritium in Molten Salt Reactors,”
Reactor Technol. 14(4)? 335 -42 (Winter 1971-72).

Discussions of the chemistry of the elements of major
significance in molten-salt reactor fuels have been made
by Grimes,’ Thoma,’ and Bae~.~ Some relevant highlights
will be summarized here.
The original fuel of the Molten Salt Reactor Experiment
consisted essentially of a mixture of ‘kiF-BeF2 -
ZrF4-UF4 (65-23-5-0.9 mole %). The fuel was circulated
at about 650°C; contacting graphite bars in the
reactor vessel and passing then through a pump and
heat exchanger. The equipment was constructed of
Hastelloy N, a Ni-Mo-Cr-Fe alloy (41-17-7-5 wt 5%).
SmaU amounts of structural elements, particularly
cliromiuna, iron, and nickel, were found in the salt.
The concentration of fission product elements in the
molten salt fuel is lower than that of constituent or
structural elements. The following estimate will indicate
the limits on the concentration of fission products
evenly distributed in the fuel.
For a single nuclide of fission yield y, at a given
power P, the number of existing nuclide atoms in the
salt is
AZFXHPX~XT
where F is the system fission rate at unit power and T IS
the effective time of operation. This is the actual time
of operation for a stable nuclide and equals l/X at
steady state for a radioactive nuclide. The contribution
to the mole fraction of the fission product nuclide in
4.5 X 1Q6 g of salt of molecular weight 40 is then
A 4.5 x lo6
x= 6X 40
As an example, for a single nuclide of 1% yield and
30-day half-life at 8 MW. A = 9.4 X IO2’ atoms and
X = 1.3 X IOs7. Because the inventory of a fission
product element involves only a few nuciides, many
radioactive. the mole fractions are typically of the order
of I x or less.
Traces of other substances may have entered the salt
in the pump bowl, where salt was brought into vigorous
contact with the purified helium cover 935. Flow of this
gas to the uff-gas system served to remove xenon and
krypton fission gases from the slytern. A slight leakage
or vxporization of oil into the pump bowl used as a
lubricant and seal for the circulating pump introduced
hydrocarbons and, by decomposition, carbon and hy-
drogen into the system. For the several times the
reactor vesse1 was opened for retrieval of surveillance
assemblies and for maintenance, the possible ingress of
cell air should be taken into account.
The binary molten fluoride system LiP-BeF2 (66-34
mole %) melts4 at about 459°C. The solution chemistry
of many substances in this solvent has been discussed
by Baes.3 Much of the redox and oxide precipitation
chemistry can be summarized in terms of lhe free
energy of formation of undissolved species.
Free energies of formation of various species calcu-
lated at 650°C largely from Baes’s data are shown in
Table 4.1. The elements of the tabk are listed in terms
of the relative redox stability of the dissolved ffuorides.
Table 4.1. Free energy of formation at 65O’C (AC’~, kcdl
Li’. Be+, and F are at uait activity; all uthers,
activities in nrole fraction units
~~ ~ ~
SoIid
- 363 36
- 344.67
-34 1.80
- 123.00
3 10.92
-389.79
-221.08
-316.93
-185.39
-2 19.42
-179.14
32.4
( 150.7)
-4.5 to 8.5
-138.18
121.58
-99.81
Dissolved in
LiFa2BeE2
126.49
-354.49
356.19
-332.14
-216.16
103.37
44.48
- 300.88
392.52
308 10
-392.92
(-296.35)
152.06
-134.59
- 11 3.40
(-186.3)
-50.29
Gas
-449.89
-366.49
306 65
259 13
-232 26
-200 59
-98 36
-42 15
-446 11
-189 57
66 12
47 04
-173 72
\./

v.&+p
As one example of the use of the free energy data, we
will calculate the dissolved CrF2 concentration suffi-
cient to hait the dissolution of chromium from Hastel-
loy if no region of lower chromium potential can be
developed as a $ink.
For the reaction
Cro(s) + 2UF4(dj = CrF,(d) t 2UFs(d),
AG = - ~K.vX - 2 [ 392.52 - (--300.88)]
= 3 1.22 kcal,
-AGO -3 1.22
log K = - - -7.39
2.3RT/lQO0 4.233
1ogK = log CrFz - log Cro 2 log (u4+/U3+) .
If we assume U~+/U~+ - 100 and note that the
chromium concentration in Hastelloy N is about 0.08
mole fraction (log Cro = -1.10).
log CrF, = 7.33 + (--I .lo>
+ 2 X 2.0 = -4.49 = log (3.2 X EQm5)
A mole fraction of 3.2 X lo-’ curresponds to a weight
concentration of 52 x 3.2 x IO-’//~O = 42 ppm Cr2+ in
solution.
To obtain a higher concentration of dissolved Cr2+,
the solution would have to be more oxidizing. Further-
more? the EIastelloy N surface during operation be-
comes depleted in chromium, and a chromium sink of
lower activity, Cr3Cz (equivalent to a mole fraction of
about 0.016 to 8.01), may be formed; all this would
require a somewhat more oxidizing regime to hold even
this much Cr*+ in solution.
The free energy data can be used to estimate the
quantities in solution only when the species shown are
dominant. Thus it is shown by Ting, Baes, and
15
Marnantov’ that under conditions of moderate concentrations
of dissolved oxide, pentavalent niobium exists
largely as an oxyfluoride, which may be stable enough
for this rather than NbF4 to be the significant dissolved
species under MSRE conditions.
The stability of the various fluorides below chromium
in the tabulation are such as to indicate that at the
rcdox potential of the U4+/U3+ couple, only the
elemental form will be present in appreciable quantity.
In particular, tellurium6 vapor is much more stable
than any of its fluoride vapors. Unless a ~ Q R
stable
species than those listed in the table exists in molten
salt, these data indicate that tellurium would exist in
the salt as a dissolved elemental gas or as a telluride ion.
[No data are available on Te2(gj; etc., but such
combinations would not much affect this view.]
References
1. W. W. Grimes, “Molten Salt Reactor Chernlstry,”
Nucl. Appl. 8, 137 55 (1970).
2. R. E. Thorna, Chemical Aspects o.f&fSRE Operutions,
ORNL-4658 (December 197 1).
3. C. F. Baes. Jr., “The Chernrstry and Thermodynamics
of Molten Salt Reactor Fuels,” Ni~l Met. 15,
6 17-44 (1 969)(USAEC COW-69080 B ).
4. K. A. Rornberger, J . Braunstein. and W. E. Thorna,
“New Electrochemical Measurements of the Liquidus in
the LiF-BeFZ System - Congruency of Liz BeF4 )” J.
P~Ys. Ci70~2. 76, 1154-59 (1972).
5. 6. Trng, C. F. Baes, Jr., and G. Mamantov, “The
Oxide Chemistry of Niobium in Molten LiF-BeF,
Mixtures,” MSR Program Semiannu. Progr. Rep. Feb.
29, 199.2, ORNL-4781, pp. 87-93.
6. Free energies for tellurium fluorides given in Table
4.1 were taken from I”. A. G. O’Ware, The Th‘hemdynaniic
Properties of Some Chalcogen Fltdorides,
ANL-73 15 (Ju~J~ 1968).

Molten-salt reactors generate the full array of fission
products in the circulating fuel. The amount of any
given nuclide is constantly changing as a result of
concurrent decay and generation by fission. Also,
certain fission product elements, particularly noble
gases, noble metals, and others, may not remain in the
salt because of limited solubility.
For the development of information on fission
product behavior from sample data, each nuclide of
each sample must be (and here has been) furnished with
a suitable basis of comparison calculated from an
appropriate model, against which the observed values
can be measured. The most useful basis is the total
inventory, which is the number of atoms of a nuclide
which are in existence at a given time as a result of all
prior fissioning and decay. It is frequently useful to
consider the salt as two parts, circulating fuel salt and
drain tank salt, which are mixed at stated times. It is
then convenient to express an inventory value as
activity per gram of circulating fuel salt, affording for
salt samples a direct comparison with observed activity
per gram of sample.
For deposits on surfaces, it is useful to calculate for
cornparison the total inventory activity divided by the
total surface area in the primary system.
Some of the comparisons for gas samples will be
based on accumulated inventory values, and others on
production rate per unit of purge gas flow. These
models will be deveioped in a later section.
In the calculation of inventory from power history,
we have in most cases found it adequate to consider the
isotope in question as being a direct product of fission,
or at most having only one significant precursor. For
the nuclides of interest, it has generally not appeared
necessary to account for production by neutron absorption
by lighter nuclides. These assumptions permit us to
calculate the amount of nuclide produced during an
interval of steady relative power and bring it forward to
a given point in real time, with unit power fission rate
and yield as factorable items.
In Table 5.1 we show yield and decay data used in
inventory calculations. In the case of Orn Ag and
34~s, neutron absorption with the stable element of
the lighter chain produced the nuclide, and special
calculations are required.
The branching fraction of l2 9Sb to 29 rnTe is a
factor in the net effective fkion yieid of Te. The
Nuclear Data Sheets are to be revised' to indicate that
this branching fraction is 0.157 (instead of the prior
literature value of 0.36). A1i our inventory values and
16
5. INVENTORY
calculations resulting from them have been proportionately
altered to reflect this revision.
The inventories for U operation were calculated
by program FISK: using a fourth-order Runge-Kutta
numerical integration method.
Differential equations describing the formation and
decay of each isotope were written, and time steps were
defined which evenly divided each period into segments
adequately shorter than the half-lives or other time
constants of the equation. The program FISK was
written in FORTRAN 3 and executed on a large-scale
digital computer at ORNL. Good agreement was ob-
tained with results from parallel integral calculations.
The FISK calculation did not take into account the
ingrowth of 239Pu during the operation with 235U
fuel. The effect is slight except for O6 RU. we obtained
values taking this into account in separate calculations
using the integral method.
For the many samples taken during operation with
233U fuel, a one- or two-element integral equation
calculation3 was made over periods of steady power.
generally nut exceeding a day. Because the plutonium
level was relatively constant (about 500 g total) during
the 3U operation, weighted yields were used, as-
suming4 that, of the fissions, 93 .% came from ' U,
2.2% from 235tJ, and 4.3% from 239Pu.
For irradiation fur an interval fl at a fission rate F
and yield Y, followed by cooling for a time t2, the
usual expressions3 for one- and twoelement chains are
Fission -+ A + B -+
where XI and X2 are decay constants for nuclides A and
B.
h program based on the above expressions was
written in BASIC and executed periodically on a
commercial time-sharing computer to provide a current
inventory basis for incoming radiochemical data from
recent samples.
To remain as current as possible. the working power
history was obtained by a daily logging of changes in

Table 5.1. Fission product data for inventory calculations
Cnmuhtive fission yieida
Chain lsutupe Half-life Fraction 234u 235u 23gh
89 Sr 52 days 1 5.86 4.79 1.711
90 Sr 28.1 years I 6.43 5.77 2.21
91 sr 9.67 hr 1 5.57 5.81 2.43
91 Y 59 days (l.0jC 5.57 5.81 2.43
35 Zr 65 days 1 .0 6.05 6.20 4.97
95 Nb 35 days (1.W 6.05 6.20 4.95
99 Mo 64 hr 1 .o 4.80 6.06 6.10
103 Ru 39.5 days 1 .0 1.80 3.00 5.64
106 WU 368 days 1.0 0.24 0.38 4.57
109 Ag Stabie (91 b + resonance) 0.044 0.030 1.40
I10 Ag(m) 253days
111 Ag 4.5 days 1 0.0242 0.0192 0.232
125 Sb 2.7 years E
0.084 0.021 0.115
127 Te(rn) 100days 0.22 0.60 0.13 0.39
129 Te(m) 34days 0.36b 2.00 0.80 2.ou
132
131
Te
I
3.25 days
8.05 davs
1 .0
1 .o
4.40
2.98
4.24
2.93
5.10
3.48
133 Cs Stable (32 b +resonance) 5.48 6.61 6.53
134 Cs 750 days
137 cs 29.9 years 1 6.58 6.15 6.63
140 Ba 12.8 days 1 5.40 4.85 5.56
141 Cc 32.3 days 1 6.49 6.40 5.01
144 Ce 284 days 1 4.41 5.62 3.93
147 Nd 11.1 days 1 1.98 2.36 2.04
€47 Pm 2.65 years 1 1.98 2.36 2.07
“From M. J. Bell, Nuclear Transmutation Data, ORIGEN Code Library; L. E. McNeese,
Engiiieering Drvebpment Studies for Molten-Salt Breeder Reactor b3-ocessirzg NO. 1,
ORNL-TM-31153, Appendix A (November 1970).
’This is the value given in the earlier literature. The revised Nuclear Data Sheets wiil
indicate that the branching fraction is 0.157.
‘Parentheses indicate nominal values.
reactor power indicated by nuclear instrumentation
charts: the history so obtained agreed adequately with
other determinations.
In practice, the daily power Bog was processed by the
computer to yield a power history which could remain
stored in the machine. A file of reactor sample times
and fill and drain times was also stored, as well as
fission product chain data. It was thus possible to
update and store the inventory of each nuclide at each
sample time. A separate file for individual samples and
their segments containing the available individual nu-
clide counts and counting dates was then processed to
give corrected nuclide data on a weight or other basis
and, using the stored inventory data, a ratio to the
appropriate reactor inventory per unit weight. The data
for individual salt and gas samples were also accumu-
lated for inclusion in a master file along with pertinent
reactor operating parameters at the time of sampling.
This file was used in preparing many tables for this
report.
Only one ad hoc adjustment was made in the
inventory calculation. Radiochemical analyses in connection
with the chemical processing of the salt to
change from &I to u fuel indicated that ~ b ,
which had continued to be produced from the 95 ZE. in
the fuel when run 14 was shut down, was entirely
removed from the salt in the reduction step in which
the salt was treated with zirconium metal and filtered.
To reflect this and provide meaningful Nb inventories
for the next several months, the calculated 95Nb
inventory was arbitrarily set at zero as of the time of
reduction. This adjustment permitted agreement between
inventory and observation during the ensuing
interval as 95Nb grew back into the salt from decay of
the Zr contained in it.
In this report we have normally tabulated the activity
of each nuciide per unit of sample as of the time of
sampling and dso tabulated the ratio of this to
inventory. For economy of space we then did not

.,. w . i
6.1 Lade Samples
Radiochemical analyses were obtained on salt samples
taken from the pump bowl beginning in run 6, using the
sampler-enricher’ J (Fig. 6.1). A tared hydrogen-fired
copper capsule (ladle, Fig. 6.2) which could contain I O
g of salt was attached to a cable and lowered by
windlass past two containment gate valves down a
slanted transfer tube unti1 it was below the surface of
the liquid within the mist shield in the pump bowl.
After an interval the capsule was raised above the latch,
until the salt froze, and then was raised into the upper
containment area and placed in a sealed transport
container and transferred to the High Radiation Level
Analytical Laboratory. Similar procedures were fol-
lowed with other types of capsules to be described
later. The various kinds of capsules had hemispherical
ends and were 3/4-in.-diarn cylinders, 6 in. or less in
length. Ladles were about 3 in. long.
After removal from the transport container in the
High Radiation Level Analytical Laboratory, the cable
CAPSULE DRIVE UNIT
LATCH-
ACCESS PORT-
AREA IC
( PRIVARY CONTAINMENT)
SAMPLE CAPSULE
OPERATIONAL AND
19
6. SALT SAMPLES
Fig. 6.1. Sampler-enricher schematic.
was slipped off, and the capsule and contents were
inspected and weighed. The top ofthe ladle was cut off.
After this, the sample in the copper ladle bottom part
was placed in a copper containment egg and agitated 45
min in a pulverizer mixer, after which the powdered salt
was transferred (Fig. 63) to a polyethylene bottle for
retention or analysis. Data from 19 such samples, from
run 6 to run 14, are shown in Tables 6.1 and 6.2 as
ratios to inventory obtained from program FESK. Full
data are given in Table 6.7, at the end of this chapter.
There will be further discussion of the results.
However, a broad overview will note that the noble-gas
daughters and salt-seeking isotopes were generally close
io inventory values, while the noble-metai group \vas
not as high: and values appeared to be more erratic.
Noblemetal nuclides were observed to be strongly
deposited on surfaces experimentaliy exposed to pump
bowl gas. This implied that some of the noblemetal
activity observed for ladle salt samples could have been
picked up in the passage through the pump bowl gas
OHNL-3IC: 63-5848R

d *@
,&-%
20
Fig. 6.2. Container f5a sampling MSRE salt.
Fig. 6.3. Apparatus for removing MSRE salt from pulverizer-mixer to polyethylene sample bottEe.

Sample Date
6-17L
6-1 9L
4-07L
7-l0L
7-12L
8-05 L
10-12L
10-2OL
11-08L
11-126,
11-22%
11-456.
ll-5BL
11-52L
11-54L
1 1-58E
12-066.
12-246.
14-22L
14-20FV
14-30FV
14-63FV
14-66FV
Sample
~- -
6-17
6-19
7-04
7-1 0
7-1 2
8-05
10-12
10-20
11-08
11-12
11-22
11-45
11-51
11-52
11-54
11-58
12-06
12-27
14-22
14-20FV
14-30FV
14-63FV
14-66FV
5-23-66
5-26-66
6-27-66
7-6-66
7-1 3-66
10-8-66
12-28-66
1-9-67
2-1 3-67
2-21-67
3-9-67
4-17-67
4-28-67
5-1-67
5-5-67
5-8-67
6-20-67
7-1 7-67
11-7-67
11-4-67
12-5-67
2-27-68
3-5-68
21
Table 6.1. Nobie-gas daughters and salt-seeking isotopes in dt samples
from MSRE pump bowl during uranium235 operation
Expressed as ratio to amount calculated for 1 g oi inventory salt at time of sampling
Noble-gas daughters Salt-seeking isotopes
SI-89 Sr-91 Sr-92 Ba-140 0-137 Ce141 (3-143 Ce-144 Nd-147 Zr-95
0.92 0.81 0.76
0.90 0.63 0.69
0.67 0.63 0.58
0.67 0.71 0.90
0.73 0.72 0.89
0.64
0.80 0.71
0.74 0.72
0.66 0.71
0.80 0.82
0.91
0.77
0.63 1.10
0.69
0.88
0.76
0.75
9.60
0.89
0.87
0.87 1.14
0.90 45.00
1.30
0.59
0.69
0.79
1.80
0.89
0.96
1.10
1.01
0.99
0.77
0.59
1.26
0.63
0.80
0.84
0.85
0.92 0.79
0.95 0.82
0.78 1.20
0.77 1.09
0.66 1.04
0.41 3.50
0.85
0.90
0.87 0.64 1.20
0.96 0.42 1 .erg
1.03 2.60 1.10
0.83 1 .06
0.84
0.76
0.77
I .20
2.40
Table 6.2. Nebte metals in salt samples from MSRE pump bowl during uranium235 opemtion
Expressed as ratios to amount calculated for 1 g of inventory salt at time of sampling
Date Kb-95
5-23-66
5-26-66
6-27-66
7-6-66
4-13-66
10-8-66
12-28-66
1-9-67
2-13-67
2-21-67
3-9-67
4-17-67
4-28-64
5-1-67
5-5-67
5-8-67
6-20-67
7-17-67
11-7-67
11-4-67
12-5-67
2-27-68
3-5-68
15.51
2.66
0.44
0.95
0.03324
0.30
0.32
0.04432
0.022 I6
0.19
0.24
0.89
0.38
0.001 11
0.00666
0.00001
0.00003
0.02216
kfo-99
~ I - - g
0.54
2.77
0.58
0.80
0.19
0.02216
0.28
1.88
1.44
1.03
0.92
0.44
0.49
0.21
0.03324
0.75
0.47
o .oi 440
0.00554
0.00222
0.00443
Ru-103
._l__li
0.01330
0.42
0.09086
Q.21
0.20
0.03767
0.02659
0.01551
0.12
0.09972
0.06648
0.21
0.05540
1.22
0.02216
0.13
0.09972
0.12
0.06648
0.00222
0,00111
0.00044
0.00033
Ru-105 Ru-106
2.50
9.31
1.33
3.74
1.44 0.29
0.06205
0.03435
0.01994
0.09972
0.09942
0.07756
0.16
0.03324
0.08864
0.02216
0.12
0.13
0.08864
0.07756
0.00665
0 .00 22 2
0.00048
0.003 3 2
Ag-111 Te-l29m
0.31
0.08089
0.12
0.17
0.09972
0.12 0.17
0.09972 0.14
0.0’3324 0.17
0.68
0.14
0.11
0.04432
0.01219
Te-l 32
0.57
0.5 1
0.44
0.40
0.33
0.14
0.17
0.47
0.31
0.42
0.31
0.14
0.14
0.088 64
(1.12
0.12
0.06648
0.00665
0.01219
0.00222
0.01219
1.12
0.95
0.95
0.71
1.09
0.98
1.09
0.23
0.86
0.96
0.94
1.10
1 .04
0.97
I .02
1.04 1.04
0.55 1.06
1.12
0.94
1-131 1-133 1-135
0.72 0.91 0.55
0.92 0.69 0.83
0.81 0.69 0.66
0.79 0.91
0.75 0.73 0.64
1.10
0.91
0.96
0.70
0.94
1 .33
1.09
0.98
0.96
0.82
0.16
0.99
8.20
0.74
0.50
0.66

and transfer tube regions, and indicated that salt
samples taken from below the surface were desirable.
4.2 Freeze-Value Samples
Beginning in run IO, gas samples (q.v.1 had been taken
using a “freeze-value” capsule (Fig. 6.4).
Tu prepare a freeze-valve capsule, it was heated
sufficiently to melt the salt seal, then cooled under
vacuum. It was thus possible to lower the capsule
nozzle below the surface of the salt in the pump bowl
before the seal melted; the vacuum then sucked in the
sample.
After the freeze-valve capsule was transferred to the
High Radiation Level Analytical Laboratory, inspected,
and weighed after removing the cable, the entry nozzle
was sealed with chemically durable wax. The capsule
exterior was then leached repeatedly with “verbocit”
VOLUME
20cc-
ORNL-DWG 61-4784A
STAINLESS STEEL
CABLE
s -3/4-in. OB NICKEL
NICKEL CAPILLARY
/--
Fig. 6.4. Freeze valve capsule.
22
(Versene, boric acid: and citric acid) and with
HNOB-HF solution until the activity of the leach
solution was acceptably low. The capsule was cut apart
in three places . in the lower sealing cavity, just above
the sealing partition, and near the top of the capsule.
Salt was extracted, and the salt and salt-encrusted
capsule parts were weighed. The metal parts were
thoroughIy leached or dissolved, as were aliquuts of the
salt.
Four samples (designated Fv were taken late in run
14 using this technique. Results shown near the bottom
of Tables 6.1 and 6.2 show that the values for
salt-seeking isotopes and daughters of noble-gas isotopes
were little changed and were near inventory, but values
for noble-metal nuclides were far below inventory. This
supports the view that the liquid salt held little of the
noble metals and that the noble-meld activity of earlier
ladle samples came from the pump bowl gas (or
gas-liquid interface) or from the transfer tube.
After run 14 was terminated the 235L fuel was
removed by fluorination, and thc carrier salt was
reduced with hydrogen and with metallic zirconium,
after which 233U fuel was added, and the system was
brought to criticality and then to power in the early
autumn of 1968.
Radiochemical analyses were obtained on ladle sam-
ples taken during treatment in the fuel storage tank
(designated FST) during chemical processing, and from
the fuel pump bowl after the salt was returned to the
fuel circulation system (designated FPj from time to
time during runs 15, 17, 1s: and 19, as shown in Table
6.3. Chemical analyses on these samples were reported
by T h~ma.~
The “Xb activity of the solution was slightly more
than accounted for in samples FPlS-GL, as the ~ir-
conium reduction process had been completed only a
short time before; the niobium inventory was set at
zero at that time. The 951\;b which then grew into the
salt from decay of 35Zr appeared in these samples to
show some response to beryllium reduction of the salt:
though this effect is seen better with freeze-valve
samples and so will nut be discussed here. The various
additions of beryllium to the fuel salt have been given
by Thoma.3
Data for all freeze-valve salt samples taken during
tj operation are summarized as ratios to inventory
salt in Tables 6.4 and 6.5, and Table 6.8 at the end of
the chapter, where various operating conditions are
given, along with the sample activity and ratio to
inventory salt. Analyses for salt constituents as well as
fission products are S ~QWII there. On the inventory-ratio
basis, comparisons can be made between any constituents
and/or fission products.

Table 6.3. Data on fuel (including car&E) salt samples from MSME pump bowl during uranium-233 operation
Ladle capsules
Values shown are the ratio of observed activity to inventory activity. both in disintegrations per minute per gram
Inventory basis, 1.4 MW = full power
SaIIlplt: Type. S-89 Sr-9 1 Y-91 Ba-140 0-137 ce-141 ce-143 0-144 ~I-147 Zr-95 Nb-95 MO-99 Mu-l 03 Ru-105 Ku-1Qh Te-129m Te-132 1-I 31
FST-25,
Aug. 14
FST-27,
Aug. 21
FST-30,
Sept. 4
FP15-6L,
Sept. 14
FPlS-9L,
Sept. 17
FPIS-1QL,
Sept. 19
FP15-1 SL,
Oct. 4
FP1526L,
Oct. IQ
FP17-IL,
Jan. 12
FPl74L,
Jan. 21
FP17-9L,
Jan. 24
FPI 7-l 2&,
Feb. 6
FPll7-l%L,
Feb. 12
FP17-191.,
Feb. 19
FP17-2QL,
Feb. 20
FP17-3QL.
Apr. 1
IS-lL,
Apr. 14
I%-5L,
Apr. 23
18-1QL,
Apr. 29
1s17L,
Aug. 27
Fuel, &we Fz
Carrier, end F2
Carrier, end Hz
Fuel
Fuel
Fuel
Fuel
Fuel
Fuel, pre power
Fuel, approach
power
FUd
Q.85 1.22 I.22 1 .Q2
0.83 I.21 0.99
0.94 1.2x 1.11
0.92 1.32 1.11
a.93 1 .Q7 1.26
0.93 1.2% 1.04
0.75 1 .Q4 0.80
0.89 1.40 1.04
(3.4) 0.84 1.32 1.05
0.85 0.37 0.82 1 .Q9
0.86 1.1.5
Fuel 1.22
Fuel
Fuel
Fuel
Fuel
Fuel
Fuel
““‘Nb removed by fluorination.
bParentheses indicate approximate value.
1.01 I.36 1.17 Cl.49 0.94 1.24 1.46 2.1x
1.12
1.17
1.01
1 .Q6
1 .Q6
1.25
1.44
0.66
0.71
0.50
(Q.26)*,@
1.30
1 .Q2
0.71
0.76
0.63
0.80
0.81
0.26
0.22
(0.11)
0.11
0.46
0.53
0.44
0.15
1.44
0.44 0.76 1.72
a.43 0.10
a.43 0.15 0.04
0.25
0.28
0.14 1.07
0.45 0.06 1.45 10.1 5.7

Tables 6.4 and 6.5 present only the ratio of the
activity of various fission products to the inventory
value for the various samples.
Two kinds of freeze-valve capsules were used. During
runs 15, 16, and 17 (except 17-32) the salt-sealed
capsule described above was used. In general, the results
obtained with this type of capsule are believed to
represent the sample fairly~ However, as discussed
above, the values for a given salt sample represent the
combination of activities of the capsule interior surface
24
with those determined for the contained sdt. To L.&i
prevent interference from activities accumulated on the
capsule exterior, as many as several dozen HN03-HF
leaching were required ; occasionally the capsule was
penetrated. Also, the salt seal appeared to leak slightly;
less vacuum inside resulted in less sample.
6.3 Double Wall C apk
When a double-walled capsule was developed for gas
samples (q.v.) it was adopted dss for salt samples. The
Table 6.4. Noble-gas daughters and salt-seeking isotopes in salt samples from MSWE pump bowl during uranium-233 operation
Sample Date
15-28
15-32
15-42
15-5 1
15-57
15-69
16-4
17-2
17-1
17-1 0
17-22
17-29
17-31
17-32
18-2
18-4
18-6
18-12
18-19
18-44
18-45
18-46
19-lb
19-6b
19-9
19-24
19-36
19-42
19-44
19-47
19-55
19-57
19-58
19-59
19-76
20-1
20-19
10-1 2-68
10-15-68
10-29-68
11-6-68
11 -1 1-68
11-25-68
12-1 6-68
1-14-69
1-2 3-69
1-28-69
2-28-69
3-26-69
4-1-69
4-3-69
4-14-69
4-1 8-69
4-23-69
5-2-69
5-9-69
5-29-69
6-1 -69
6-1-69
8-1 1-69
8-15-69
8-18-69
9-10-69
9-29-69
10-3-69
10-6-69
10-7-69
10-14-69
10-17-69
10-17-69
10-19-69
10-30-69
11-26-69
12-9-69
"Approximate value.
hFlush salt.
Expressed as ratio to amount calculated for 1 g of inventory salt at time of sampling
Noble-gas daughters Salt-seeking Isotope$
__ -.
SI-89
-
SI-90 Y-91 Ba-140 Cs-137 (3-141 (3-144 Nd-147 Zr-95
0.20
0.94
0.84
0.79
0.87
1.01
0.69
0.48
0.60
0.55
0.77
0.63
0.46
0.77
0.78
0.60
0.97
0.62
0.76
0.75
0.72
0.00863
0.01 677
0.70
0.89
0.82
0.86
0.48
0.89
0.74
0.75
0.73
0.65
0.73
0.63
0.53
0.61
0.70
0.89 1.24
0.71 0.66
0.11 Q.96
0.76 1.22
1.31
1.22
2.53 0.64
0.94
1.90
1.04
1.34
1.38
1.11
1.12
1.02
0.00389
0.01612
0.39
I .37
I .00
1.12
1.21
1.23
1 .00
1.13
1.17
2.23
0.76
I .29
1.02
1.17
(1.42)"
0.80
0.69
0.38
1.08
1.05
1 .O1
0.92
1.10
0.81
I .I4
1.01
1.06
1.04
0.00464
0.02 I57
0.85
(!.I9
0.94
0.98
0.99
0.98
1.12
1.10
1.03
1.10
I .04
1.07
0.91
1.36
0.84
0.82
0.81
0.93
0.84
0.69
0.44
0.68
0.83
0.0977 6
0.91
0.81
0.77
0.84
0.88
0.81
0.72
0.92
0.79
0.91
0.84
0.02514
0.099Ki
0.52
0.85
1.10
1.36
0.08 140
0.80
0.69
0.85
0.86
0.80
0.98
0.75
0.79
0.5 3
0.94
0.80
8.85
0.77
0.80
0.91
0.78
0.71
0.81
0.81
0.78
0.81
0.75
0.00176
0.15
0.60
0.75
0.76
0.84
0.82
0.86
0.90
0.87
0.81
0.79
0.85
0.67
0.69
0.29
1.08
1.16
1.13
I .25
I .06
1.10
1.10
0.70
0.62
1 .07
1.28
1.11
1.16
1.22
1.21
1.03
1.23
1.10
0.94
1 si2
1.23
0.00984
0.02149
0.91
1 .OO
1.07
1.12
1.09
1.15
1 .22
1.16
1.17
1.06
1.16
1.13
1 .Oh
0.66
0.98
0.99
1.22
1 .08
1.16
1.49
1.15
0.65
0.04 I 48
1.20
1 .O%
0.014 39
0.18
0.12
0.94
1.21
1.18
1.20
1.34
I .22
1.36
1.16
1.19
0.28
0.91
0.88
1.14
0.98
0.9 3
1.38
0.79
0.72
0.89
0.89
0.98
0.92
0.96
0.97
0.99
0.99
0.94
I .01
0.55
0.95
0.90
0'00286
0.0101s
0.42
0.86
0.86
0.93
1.01
0.99
0.95
0.9 3
0.92
0.90
0.74
0.92
0.85

.-
interior copper capsule was removed without contact
with contaminated hot-cell objects and was entirely
dissolved. The outer capsule could also be dissolved to
determine the relative amounts of activity deposited on
such a “dipped specimen.” Data on capsule exteriors
will be given in a separate section. Salt samples
beginning with sample 14-32 were obtained using the
25
double-walled capsule. Operating conditions associated
with the respective san~ples are summarized in Table
6.6.
We should note that very little power had been
produced from 233W prior to sample 15-63; much of
the activity was carried over from 235U operations.
Sample 17-2 was taken during the first approach to
Table 6.5. Noble metals in salt samples from MSRE pump bowl during uranium-233 operation
Expressed as ratio to amount calculated fur 1 g of inventory salt at time of sampling
Sample Date Nb-95 hlo-99 Ru-103 Ru-106 Ag-1 I 1 Sb-125 Te-129 m Te-132 1-1 3 I
15-28
15-32
15-42
15-51
15-57
15-69
16-4
17-2
17-7
17-1 0
17-22
17-29
17-31
17-32
18-2
18-4
18-6
18-1 2
18-19
18-44
1845
18-46
19-lC
19-6‘
19-9
19-24
19-36
1942
19-44
19-47
19-55
19-57
19-5 8
14-59
19-76
20-1
20-19
10-12-68
10-1 5-68
10-29-68
11-6-68
11-1 1-68
11-25-68
12-16-68
1-1 4-69
1-23-69
1-28-69
2-28-69
3-26-69
4-1-69
4-3-69
4-14-69
4-1 8-69
4-23-69
5-2-69
5-9-69
5-29-69
6-1-69
6-1-69
8-1 1-69
8-1 5-69
8-1 8-69
9-10-69
9-29-69
10-3-69
10-6-69
10-7-69
10-1 4-69
10-1 7-69
10-17-69
10-17-69
10-30-69
11-24-69
12-5-69
0.74
1.16
0.42
0.85
1 .Q6
0.02000
0.54
0.52
0.29
-0.02312’
-0.05000
0.51
0.32
0.31
0.46
0.22
1.56
0.04847
0.01641
-0.03579
0.44
0.02242
(0.11)
(-0.00108)
0.34
0.33
-0.08623
-0.06114
- 0.05268
0.02630
5.05050
-0.03366
-0.01 449
0.02995
0.19
0.05389
0.09635
(2.48P
0.12
0.53
0.00971
0.00445
0.01360
0.00344
0.01496
0.00839
0.85
0.00812
0.00773
0.03459
1.36
0.12
0.22
0.01916
0.43
0.41
0.83
0.63
0.75
0.01885
0.19
0.01412
3.34
0.78
0.02536
0.00006
0.00096
0.00037
0.00433
0.01241
0.04953
0.03352
0.01 134
0.00076
0.02199
0.00 I77
0.0221 6
0.52
0.07401
0.00160
0.00202
0 .00 I5 3
0.00862
(0.07066)
(0.00308)
0.00071
0.21
0.0021 9
0.1 3
0.04484
0.15
0.18
0.28
0.00367
0.03270
0.00195
0.29
0.11
0.03436
0.000 15
0.00084
0.00026
0.00329
0.00442
0.00304
0.00165
0.00055
0.000 2 7
0.23
0.02958
0.00327
(0.02304)
(0.00192)
0.23
0.08871
0.51 308
0.05326
0.05759
0.1 1
0.00207
0.01478
0.14
0.05074
(2.57)
0.09095
0.00425
0.1 3
0.08057
0.03367
1.23
0.06622
0.05940
5.13
0.03976
8.1 2
0.20
0.60
0.10
0.04055
Q.48
0.23
0.14
0.00007
0.00391
o.ooaii
0.00195
0.00004
0.00687
0.06829
0.26
0.21
0.01981
0.00233
0.01014
0 .o 1 9 2 1
1.83
0.00383
0.05106
0.02532
(0.10)
(0.01 185)
0.74
0.00335
0.18
0.19
0.13
0.30
0.02334
0.22
0.28
0.09176
(3.67)
0.31
0.03021
0.02040
0.01 169
0.03794
0.00348
0.006 7 L!
0.00398
I .80
0.00230
0.0081 3
0.14
0.04 30 2
0.24
o .a061 4
0.09806
0.01808
0.06358
0.04204
0.07974
0.0015 1
0.00680
8.00031
0.55
0.21
1.13
(1.67)
0.40
0.37
0.46
0.28
0.30
0.40
0.10
0.35
0.26
0.53
0.74
0.34
0.44
0.08227
(0.28)
“Parentheses indicate approximate value.
%egatlve nunntms result when 95Nb, which grows in from 95Zr present between sampling and analysis time, exceeds that found
by analysis.
‘Flush salt.
0.60
0.58
0.65
0.89
0.11
0.44
0.54
0.64
0.15
0.40
0.68
0.41
i

Sample
NO.
15-28
15-32
15-42
15-5 z
15-57
15-69
16-4
17-2
17-7
17-10
17-22
17-29
17-3 1
17-32
18-2
18-4
18-6
18-12
18-19
18-44
18-45
18-46
19-la
19-6a
19-9
19-24
19-36
19-42
19-44
19-47
19-55
19-57
19-58
19-59
19-76
20-1
20-1 9
‘Flush salt
Equivalent
Date The full-power
hours
10-12-68 1726
10-15-68 2047
10-29-68 1123
11-6-68 1531
11-1 1-68 2145
11-25-68 1700
12-16-68 0555
1-14-69 1025
1-23-69 1320
1-28-69 0603
2-28-69 2259
3-26-69 1506
4-1-69 1145
4-3-69 0552
4-14-69 1150
4-18-69 2119
4-23-69 1015
5-2-69 1305
5-9-69 1925
5-29-69 0311
6-1-69 0921
6-1-69 1412
8-1 1-69 0845
8-15-69 0413
8-18-69 0604
9-10-69 1049
9-29-69 1108
10-3-69 1105
10-6-69 0635
10-7-69 1033
10-14-69 1047
10-17-69 0620
10-17-69 094 1
10-17-69 1240
10-30-69 1159
11 -26-69 1704
12-5-69 0557
Table 6.6. Operating conditions for salt samples taken from MSRE pump bowl during uranium-233 operation
0.01
0.01
0.01
0.01
0.01.
0.01
1.56
1.69
94.00
155.50
719.63
1144.75
1271.00
1307.00
1527.75
1601.25
1713.12
1939.00
2106.00
2473.00
2538.63
2538.63
2538.63
2538.63
2538.63
2781.87
2978.50
3048.87
3118.62
3148.75
3330.25
3395.38
3397.13
3397.13
3705.63
3789.38
3990.87
Overflow
.___
Percent of Hours at
Purge gas flow
Pump Voids Pump bowl Previous
full percent of level
rpm (72)
Lb/hr return
power power
- Std Gas Psig
(%) liters/rnm
Day Time
.~
0.00
0.01
0.00
0.00
0.00
0.00
0.00
5.63
57.50
58.75
87.50
86.25
90.00
90.00
100 .00
100.00
100.00
100.00
100.00
86.25
0.00
0.00
0.00
0.00
0.13
0.13
68.75
8’7.50
100 .00
100.00
100.00
100.00
0.13
0.13
100.00
100.00
100.00
0.5
0.1
9.4
24.3
0.5
21.4
62.8
1.5
15.1
39.0
31.4
0.1
140.8
182.9
41.8
49.8
158.7
376.5
93.8
28.7
0.4
5.2
0.0
0.0
1.1
19.6
66.1
49.0
63.3
91.3
259.5
41.6
1 .0
4.0
140.6
1 .7
36.7
1180
1180
1180
1180
1180
1180
1180
1180
1180
1180
942
1050
1050
1050
1180
1180
1180
1180
11 80
990
990
990
1189
1170
1189
1165
608
1176
1188
1175
1186
1186
1188
1189
1176
1190
1200
0.00
0.60
0.00
0.00
0.60
0.60
0.60
0.60
0.60
0.60
0.00
0.05
0.05
0.05
0.60
0.60
0.60
0.60
0.60
0.00
0.00
0.00
0.00
0.00
0.60
0.70
0.00
0.53
0.5 3
0.53
0.53
0.53
0.53
0.53
0.53
0.53
0.53
63
66
66
66
63
56
62
59
57
57
65
58
62
6 0
61
63
63
59
59
53
50
56
72
62
65
62
58
68
64
61
63
63
67
65
66
64
63
1.4
2.9
3.7
1 .o
0.8
0.8
0.8
3.8
1.8
1.6
0.7
1.3
3.0
4.5
7.4
4.9
4.7
3.0
0.9
0.0
0.0
0.0
0.0
0.7
2.4
4.7
0.9
6.3
7.4
1.8
2.6
3.7
9.0
3.9
8.6
6.8
5.6
10-12
10-15
10-28
114
11-11
11-25
12-15
1-14
1-23
1-27
2-27
3-25
4-1
4-2
4-14
4-18
4-23
5 -2
5-9
5-28
5-3 1
5-3 I
00
00
8-18
9-10
9-27
10-3
10-3
10-7
10-14
10-17
10-17
10-17
10-30
11-26
12-5
0711
1830
1730
1630
1208
04 25
1758
0310
07 36
2315
1630
1132
1015
1807
0853
1901
0733
0500
1303
1833
2223
2223
0 0
0 0
0219
0501
0316
1036
0305
0228
0353
0105
0757
0757
09 16
1320
0248
3.30
3.30
3.30
3.30
3.30
3.30
3.30
3.30
3 30
3.30
3.30
3.30
3.30
3.30
3.30
3.30
3.30
3.30
3.30
2.30
2.00
2.00
3.30
3.30
3.30
2.90
3.35
3.30
3.30
3.35
3.30
3.30
3.30
3.30
3.30
3.30
3.30
He
He
He
IIe
He
He
He
He
He
He
He
He
He
He
IIe
He
He
He
He
He
Ar
AI
He
He
He
Ar
€He
He
He
He
1% e
He
ne
He
He
He
He
5.2
5.5
4.9
5.0
5.5
5.0
4.6
3.8
4.2
5 .0
5.4
4.2
8.9
3.2
4.6
5.2
5.5
5.2
4.8
13.0
12.8
13.6
2.4
5.3
5.3
5.0
6.3
5 .0
5.5
5.8
5.2
5.2
5.6
5.6
5.2
5.5
5.2
Purge on
Purge
Purge on
Purge on
Purge on
Purge on
Purge on
Purge on
Purge on
Purge on
Purge on
Purge on
Purge on
Purge on
Purge on
Purge on cm
Purge on
Purge on
Purge on
Purge on
Purge on
Purge on
Purge on
Purge on
Purge on
Purge off
Purge off
Purge off
Purge off
Purge off
Purge off
Purge off
Purgc off
Purge off
Purge off
Purge off
Purge off
h)

sustained high power; the higher values for the shortestlived
nuclides reflect some uncertainties in inventory
because heat-balance calibrations of the current power
level had not been accuniplished at the time - it
appeared more desirable to accept the inventory aberration,
significant only for this sample, than to guess at
correction.
Sample 18-46 was taken 5'4 hr after a scheduled
reactor shutdown. Samples 19-1 and 19-6 are samples
of flush salt circulated prior to returning fuel after the
shu tdown ~
6.4 Fission Product Element Grouping
It is useful in examining the data from salt samples to
establish two broad categories: the salt-seeking elements
and the noble-metal elements. The fluorides of thc
salt-seeking elements (Wb, Sr, Y, Zr, Cs, Ba, La, Ce, and
rare earthsj are stable and soluble in fuel salt. Some of
these elements (Wb, Sr, Y, Cs, Ba) have noble-gas
precursors with half-lives long enough for some of the
noble gas to leave the salt before decay.
Noble-metal fission product elements (Nb, Mo, Tc,
Ru, Rh, Pd, Ag [Cd: In, Sn?], Sb, Te, and I) do not
form fluorides which are stable in salt at the redox
potential of the fuel salt. Niobium is borderline and will
be discussed later. Iodine can form iodides and remain
in the salt; it is included with the noble nietals because
most iodine nuclides have a tellurium precursor - and
also to avoid creating a special category just for iodine.
The Nb-Mo-Tc-Ru-Rh-Pd-Ag elements for a subgroup,
and the Sb-Te-I elements another.
Quite generally in the salt samples the salt-seeking
elements are found with values of the ratio to inventory
activity not far from unity. Values for some nuclides
could be affected by loss of noble-gas precursors. These
include:
RecurmoP NMdide affected
3.18-min ""Kr S'Sr
33-sec "Kr 90%
9.8-sec 9' Kr
3.9-min 'j7xe
9ISr, 9'Y
13 7cs
1B-sec * 40~e 140Ba
The "Sr and IS7Cs in particular might be expected
to be stripped to some extent into the pump bowl gas,
as discussed below for gas samples. In Table 6.4, ratio
values for 'lY and '"Ba are close to unity and
actually slightly above.
Values for
Ce run slightly below unity, and those
for 144Ce (and 147Nd) somewhat above. The 95Zr
values average near but a few percent below unity.
Thus the group of salt-seeking elements offers no
surprises, and it appears acceptable to regard them as
27
remaining in the salt except as their noble-gas pre-
cursors may escape.
The general consistency of the ratio values for this
group provides a strong argument for the adequacy of
the various channels of information which come tu-
gether in these numbers: sampling techniques, radio-
chemical procedures, operating histories, fission prod-
uct yield and decay data, and inventory calculations.
6.5 Noble-Metal Behavior
The consideration of noble-metal behavior is ap-
proached from a different point of view than for the
salt-seeking group. Thermodynamic arguments indicate
that the fluorides of the noble metals generally are not
stable in salt at the redox potential of MSWE opera-
tions. Niobium is borderline, and iodine can form
iodides, which could remain in solution. So the ques-
tiopls are: Where do the noblemetal nuclides go, how
long do they remain in salt after their formation before
leaving, and if our salt samples have concentrations
evidently exceeding such a steady state, how do we
explain it? The ratio of concentration to inventory is
still a good measure of relative behavior as long as OUT
focus is an events in the salt.
If the fluorides of noble-metal fission product ele-
ments are not stable, the insolubility of reduced
(metallic or carbide) species makes any extra material
found in solution have to be some sort of solid
substance, presumably finely divided. Niobium and
iodine - later tellurium - will be discussed separately,
as these arguments do not apply at one point or
another.
If we examine the data in Table 6.5 for Mo, Ru, Ag,
Sb, and Te isotopes during runs 15 to 20, it is evident
that a low fraction of inventory was in the salt. We
simply need to decide whether what we see is dissolved
steady-state material or entrained colloidal particulate
material.
If the dissolved steady-state concentration of a
soluble material is low, relative to inventory, loss
processes appreciably more rapid than decay must exist.
If the average power during the shorter period required
to establish the steady state is fi ~ then at steady state it
may be shown that
It follows that the ratio of observed to inventory
activity will be:
all periods

The amounts in solution should be proportional to the
inverse of half-life, to the current power (vs fulI), and
to the relative degree of full-power saturation. The
amount does not depend on the other atoms of the
species as long as the loss term is first order.
It follows that samples taken at low power after
operation at appreciable power should drop sharply in
value compared with the prior samples. These include
18-45, 18-46, 19-24, 19-58, and 19-59. Of these
samples, only 19-58 appears evidently low across the
board; the criterion is not generally met.
The expression also indicates that after a long
shutdown, the rise in inventory occurring (for a half-life
or so), with fairly steady loss rate? should result in an
appreciable decrease in the ratio. The beginnings of runs
1'7 and I9 are the only such periods available. Here the
data are too scattered to be conclusive; some of the
data on '*"Te and 132Te appear to fir: samples 17-7
and 19-24, respectively, are somewhat higher than
many subsequent samples.
Brigs4 has indicated that the loss coefficients should
be
for mass transfer to graphite, metai. and bubble surface,
respectively, if sticking factors were unity and K the
ratio of the mass transfer coefficient to that of xenon.
For metal atoms, K - 1; neglecting bubbles, I, - 7
hr-' .
The ratio to inventory predicted above is dominated
by the first factor, X/(h + L), in the cases (a majority)
where the present power was comparable with the
average power for the last half-life or so. We can then
note for the various nuclides using L = 7 hr-' :
Nuclide Half-life (days) hi(r, +
""h10 2 79 0 0015
IQ3RU 39 6 0 OQOl0
'06RU 361 0 OOOOI
111
Ag 75 0.00055
Comparing the observed ratios with these shows that
what we observed in essentially ail cases was an order of
magnitude or more greater. This indicates that the
observed data have to be accounted for by something
other than just the steady-state dissolved-atom concen-
tration serving to drive the mass transfer processes.
The concept that remains is that some form of
suspended material contributed the major part of the
activity found in the sample. Because this would
represent a separate phase from the sait, the mixture
28
proportion could vary. The possible sources and behavior
of such a mixture will be considered in a later
section after other data, for surfaces, etc.. have been
presented. We believe that the data on noblemetal
fission products in salt are for the major part explicable
in terms of this concept.
Three elements included in the table of noblemetal
data should be considered separately: niobium, iodine,
and tellurium.
6.6 Niobium
Our information on niobium comes from the 35-day
9s Nb daughter of 65-day " Zr. Thermodynamic con-
siderations given earlier indicate that at fission product
concentration levels, Nb4+ is likely to be in equilibrium
with niobium metal if the redox potential of the salt is
set by U3+/U4+ concentration ratios perhaps between
0.81 arid 0.001. If Nb3+ species existed significantly at
MSRE oxide concentrations, the stability of the soIuble
form would be enhanced. If NhC were formed at a rate
high enough to affect equilibrium behavior, then the
indicated concentration of soluble niobium in equi-
Iibriuin with a solid phase would be considerably
decreased.
Because "Nb is to be considered as a soluble species,
direct comparison with inventory is relevant in the
soluble case (when insoluble? it should exhibit a limiting
ratio comparable to 34-day * mTe, or about 0.0001).
The data for 95"b in salt sarnpIes do appear to have
substantial ratio values, generally 0.5 to 0.3 at times
when the salt was believed to be relatively oxidizing.
When appreciable amounts of reducing agent, usudly
beryllium metal, had been added, the activity relative to
inventory approached zero (50.05). Frequentiy, slightly
negative values resulted from the subtraction from the
observed niobium activity at count time of that which
would have accrued from the decay of "23 in the
sample between the time of sampling and the time of
counting.
6.7 Iodine
+.-
Iodine, exemplified by 311, is indicated to be in the
form of iodide ion at the redox potential of fuel salt,
with little I2 being stripped as gas in the pump bowl.5
Thermodjyiarnic calculations indicate that to strip 0.1%
of the ' I as I2 ~ a U4+/IJ3+ of at least lo4 would have
to exist. As far as is known, the major part of MSRE
operation was not as oxidizing as this (however, because
surne dissociation to iodine atoms can occur in the
vapor, stripping could be somewhat easier). I ~+&&

.*.e activities relative to inventory were between 8 and
113%, with most vaiues falling between 30 and 60%.
What happened to the remainder is of interest. It
appears likely that the tellurium precursor (largely
25-min 31Te) was taken from solution in the salt
before half had decayed to iodine, and of this tellurium,
some, possibly half, might have been stripped: and the
remainder deposited on surfaces. Perhaps half of the
"'I resulting from decay should recoil into the
adjacent salt. From such an argument we should expect
about the levels of I that were seen.
6.8 Tellurium
Tellurium is both important and to some extent
unique among the noble metals in that the element has
a vapor pressure at reactor temperature (6SO"C) of
about 13 torr. Since the fluorides of tellurium are
unstable with respect to the element, at the redox
potential of the fuel, we conclude that the gaseous
element and the tellurium ion are the fundamental
species. As a dissolved gas its behavior should be like
xenon. The mass transfer loss rate coefficients indicated
by Briggs4 would apply, E - (12X + S.7K + 7Kj, so
that with high sticking factors, about -hr production
would be the steady-state concentration, and half the
tellurium would go to off-gas. The dissolved concentration
for "Te, relative to inventory salt, would be
about 0.0006, and for '"Te, about Q.OOOQ6. Again it
is evident that the observations run higher than this.
Recent observations by @. E. Bamberger and J. P.
Young of ORNL suggest that a solubie, reactive form of
telluride ion can exist in molten salt at a presently
undefined redox potential. Such an ion could be an
important factor in tellurium behavior. However, it is
also plausible that tellurium is largely associated with
undissolved solids, by chemisorption or reaction. Any
of these phenomena would result in lower passage as a
gas to Off-jids.
The viewpoint that emerges with respect to noble-
metal behavior in salt is that what we see is due to the
appearance of highly dispersed but undissolved materid
in the salt, a mobile separate phase, presumably solid,
which bears much higher noble-metal fission product
concentrations than the salt. Our samples taken from
the pump bowl can only provide direct evidence
concerning the salt within the spiral shield, but if the
dispersion is fine enough and turnover not too slow, it
should represent the salt of the pump bowl and
circulating loop adequately.
We have suggested that the noble metals have behaved
as a mob& separate phase which is concentrated in
29
noble metals and is found in varied amounts in the salt
as sampled. This is illustrated in Fig. 6.5, where the
activities of the respective nuclides (relative to inven-
tory salt) are plotted logarithmically from sample to
sampie. Lines for each nuclide from sample to sample
have been drawn. A mobile phase such as we postu-
lated, concentrated in the noble metals, added in
varying amounts to a salt depleted in noble metals,
should result in lines between samples sloping all in the
same direction. Random behavior wouid nut result.
Thus the noblemetal fission products do exhibit a
common behavior in salt, which can be associated with
a commm mobile phase.
The nature and amount of the mobile phase are nut
established with certainty, but several possibilities exist,
including (1) graphite particles, (2) tars from decom-
posed lubricating oil from the pump shaft, (3) insoluble
colloidal structural metal in the salt: (4) agglomerates of
fission products on pump bowl surface and/or bubbles,
(5) spalled fragments of fission product deposits on
graphite or metal. As we shall later see, at least some of
the material deposits on surfaces, and it is also indicated
that some is associated with the gas-liquid interface in
the primp bowl.
References
1. R. B. Gallaher, Operation of the Sampier-Enricher
in the Molten Salt Reactor Experimen f, OR?&-TM-
3524 (October 1391).
2. R. C. Robertson, MSRE Design arid Operation
Report, Part 1. Description of Reactor Design.
ORNL-TM-728 (January 1965).
3. R. E. Thoma, Chemical Aspects of ABRE Operafi(9~1,
8RNL-4658 (December 197 I).
4. R. B. Briggs and J. R. Tallackson, "Distribution of
Noble-Metai Fission Products and Their Decay Heat ,"
MSR Program Semiannu. Progr. Rep. Feb. 28, 1969,
ORNL-4396, pp. 62-64; also a. B. Briggs, "Estimate of
the Afterheat by Decay of Koble Metals in MSBR and
Comparison with Data from the MSRE." internal
ORNL memorandum, November 1968. (Internal document
no dissemination authorized.)
5. See R. P. Wichner, with C. F. Baes, "Side Stream
Processing for Continuous Iodine and Xenon Removal
from the MSWK Fuel," internal memorandum BRNL-
CF-72-61? (Jnne 30, 1972). (Internal document no
further disserninatioli authorized.)

11VS htlOlN3ANI 01 3AIlVl3tl AllAll3V
6
L -
6
4
R

t p
ul
c
d
N P.
2
d
Gi
c??
dd
I:

Sample number 15-28s
Sample weight, g 1.3328
Date 10-12-68
Megawatt-hours 0.1
Power, hrw 0.00
RPm 1180
Pump bowl level, % 63.20
Overflow rate, b/hr 1.4
Voids, % 0.00
Flow rate ofgas, std liters/min 3.30 He
Sample line purge On
SI-89 52.00 5.46 1.32B9
0.198
SI-90 10.264.00 5.86
Y-9 1 58.80 5.57
Ba-140 12.80 5.40
Cs-137 10,958.00 6.58 5.5389
1.355
Ce-144 284.00 4.61 1.43E10
0.295
Zr-95 65.00 6.05 4.04B9
0.279
Nb-95 35.00 6.05 6.39E9
0.740
Mo-99 2.79 4.80
Ru-103 39.60 1.99 5.1387
0.025
Ru-106 367.00 0.43 1.9788
0.034
Sb-125 986.00 0.08 1.62E7
0.143
Te-132 3.25 4.40
I131 8.05 2.90
Constituent
U-233
U, Total
Li
%e
Zr
32
Table 6.8. Data for salt samples from MSRE pump bowl during uranium-233 operation
1.222
0.151
44.35
0.384
25.28
0.378
41.3
0.357
15-323 1542s 15-51s
60.9115 51.5684 56.9563
10-15-68 10-29-68 11-6-68
0.1 0.1 0.1
0.00 0.00 0.00
1180 1180 1180
65.50 66.50 66.00
2.9 3.7 1 .0
0.60 0.00 0.00
3.30 He 3.30 He 3.30 He
On On On
5.9889
0.936
2.4889
0.609
3.42B9
0.840
5.18E10
1.077
1.28E10
0.914
1.04E10
1.159
l.lSE5
0.000
8.6E5
0.000
7.48E3
0.000
7.86
0.974
63.3
0.948
109.1
0.949
Fission Product isotope@
15-57s
32.0493
11-1 1-68
0.1
0.00
1 180
62.50
0.8
0.60
3.30 He
On
4.4789 3.7689 3.8489
0.839 0.787 0.871
2.8689
0.703
3.35E9
0.823
5.4 1 E10
1.163
1.07E10
0.884
7.12E9
0.718
3.3089
0.81 1
5.17E10
1.134
1.26E10
1.135
8.7189
0.854
3.77E9
0.926
5.61E10
1.249
1.02E10
0.981
1.08810
1.059
1.4586 4.7785 5.1286
0.001 0.000 0.004
4.6586 1.4286 1.7887
0.001 0.000 0.003
4.38E5 1.18E4 2.14B5
0.004 0.000 0.002
7.50
0.929
109.8
0.95 1
57.9
0.867
109.9
0.907
15-69s
51.9611
11-25-68
0.1
0.00
1180
56.00
0.8
0.60
3.30 He
On
3.4289
0.842
4.61E10
1.060
8.4689
0.935
2.0288
0.020
4.6483
0.000
7.82
0.969
68.3
1.022
105.3
0.910
164-FVS
56.0914
12-16-68
12.5
0.00
1180
62.30
0.8
0.60
3.30 He
On
2.9889
1.014
3.63E9
0.894
6.4489
1.236
2.5 1E8
1.167
2.81B9
0.692
4.55E10
1.099
1.02810
1.382
5.0889
0.537
3.1786
0.096
8.7586
0.012
2.2587
0.004
7.4285
0.007
3.9086
0.092
1.1488
1.129
6.53
0.977
7.85
0.973
108.0
0.942
65.1
0.975
87.5
0.756

Constituent
U-2 3 3 7.28
1 .089
U-total 23.54
2.917
Li 116.4
1.008
Be 65.97
Q.988
2r 138.4
1.196
7.11
1.044
6.18
0.766
107.5
0.931
65.1
0.945
135.1
1.168
33
Table 6.8 (continued)
Samde number 18-2-NFVS 184-NFVS 186-NFVS 18-12-NFV 18-19-NFV 1844-NFV 1845-NFV 1844-NFV
Sample weight, g
3.8651
Date
4-14-69
Megawatt-hours
12,222.0
Powm, kiw
8.00
Mm
I180
hmp bowl kVd, % 61.30
Overflow rate, Ib/h 7.4
Voids, %
0.60
Flow rate of gas, std liters/min 3.30 He
Sample Line purge On
11.1866
4-18-69
12,810.0
8.00
1180
63.10
4.9
0.60
3.30 He
On
6.7612
4-23-69
13,705 .O
8.00
1180
63.20
4.7
0.60
3.30 He
On
12.6825
5-269
15,5 12.0
8.00
1180
59.50
3 .0
0.60
3.30 He
On
12.669 f
5-969
16,848.0
8.00
1180
58.90
0.9
0.60
3.30 He
OR
3.6339
5-2969
19,784.0
6.90
990
52.90
0.0
0.00
2.30 He
On
14.0790
6-169
20,309.0
0.00
990
50.00
0.0
0.00
2.00 Ar
OR
14.9780
6-1-69
20,309.0
0.00
990
55.30
0.0
0.00
2.00 Ar
OK
Fission product isotope@
Half-life
Fission
yield
Isotope (9%)
SI-89 52.00 5.46 7.36E10 7.60B10 4.24E10 1.12Ell 7.516310 1.00E1 I 1.01Ell 9.57ElO
0.771 0.780 0.600 0.966 0.621 0.758 Q.754 0.720
Y-9 1 58.80 5.57 1.44E11 9.02E10
1.38Ell 1.49E1 1 1.32E11 1.35Ell 1.23Ell
1.702 1.043
1.340 1.380 1.109 1.116 1.019
Ba-140 12.80 5.40 1.21Ell 1.43Ell 1.17Ell
1.84EE 1 1.56Ell 1.65Ell 1.48E11
0.911 1.100 0.813
1.143 1 . OM 1.058 1.039
cs-137 10,958.00 6.58 4.06E9 4.30E9 4.03E9 3.66h9 4.74E9 4.25E9 4.98E9 4.58B9
0.835 0.878 0.811 0.718 0.915 0.786 0.914 0.840
0-141 33.00 7.09 1.29Ell 1.12Ell 1.09E11 1.38Ell 1.41E11 1.45Ell 1.52Ell 1.39Ell
0.908 0.783 0.712 0.807 0.806 0.784 0.813 0.747
Ce-144 284.00 4.61 6.28E10 6.3OE10 5.50ElO 6.82EI0 6.25E10 5.67ElO 6.2i~ia 7.54~10
1.222 1.214 1.034 1.227 1.096 0.937 1.016 1.234
Nd-147 11.10 1.98 9.29E10 5.52E10
3.90E10 2.36E9 6.82B10 S.80ElO
1.488 1.150
0.653 0.041 1.196 1.030
ZI-95 65 .OO 6.05 8.50E10 8.96E10 7.40EIO 1.01 EI 1 1.13Ell 1.18E11 1.20Ell 1.13Ell
m-95 35 .00 6.05
0.966
2.40E10
0.994
1.23E10
0.792
9.09~1 a
0.944
3.17E9
1.009
1.17E9
0.952
-3. 1OE9
0.952
3.33E10
0.904
- 2.00E9
0.462 0.224 1.562 0.048 0.016 -0.036 0.374 -0.022
MO-99 2.79 4.80 1.75E9 9.99B8 1 .?BE1 B 1.38~9 1.16E9 4.22E9 1.8OEll 1.45E10
0.015 0.008 0.851 0.008 0.008 0.035 1.364 0.116
Ru-103 39.60 1.99 E.93E10
2.97E9 7.13E7 9.30E7 7.58E7
4.29E8
0.521
0.074 0.002 0.002 0.002
0.009
Ru-106 367.00 0.43 1.28E9
1.68E8
2.03E7
0.230
0.030
0.003
Ag-111 7.50 0.02 2.02E7
0.034
8.35M
1.232
4.90E7
0.066
4.03E7
0.060
8.76E7
Q.128
2.44E4
0.040
Sb-125 986.00 0.08
1.68E7
0.068
Te-129m 34.00 0.33
1.29E10 3.02E7
4.40E8 2.17B8
1.828 0.004
0.051 0.025
Te-132 3.25 4.40 7.24E8 4.30B8 2.49E11 3.5638
9.27638 1.74ElB 4.99E9
0.007 0.004 1.804 0.002
0.088 0.144 0.043
1-131 8.05 2.90 7.33E9 2.49E10 2.11EIO 5.56310 6.59E10 2.78E10 3.62ElO 6.68E9
0.101 0.352 2.260 0.595 0.739 0.339 0.438 0.082
Salt constic tuentsb
6.233
0.933
6.936
0.859
64.34
Q.557
60.93
0.912
91.4
0.790
6.72
1 .005
98.56
0.853
7.19
1.076
6.87
0.851
110.7
0.958
70.0
1 .W8
103.9
0.898
6.88
1.029
6.96
0.862
103.2
0.894
62.5
0.936
128.3
1.109
7.10 6.58
1.062 0.984
8.88 6.75
1.100 0.836
106.7 98.6
0.924 0.854
64.0 40.6
0.998 1.057
110.8 111.7
0.958 0.965

34
TabL 6.8 (continued)
Sample RiilllbeF 19-55-FVS 19-57-FVS 19-58-FVS 19-59-FVS 19-76FVS 20-1 -FVS 20-19-FVS
Sample WRkht, g 4.2936 3.3515 12.3829 8.4980 13.4771 3.1065 5.4784
Date 10-1469 IO-17-69 10-1769 10-1749 10-3069 11-2669 12-5-69
Megawatt-hours 26,642.0 27,163.0 27,177.0 27,177.0 29,645.0 30,315.0 31,927.0
Power, RIW 8.00 0.00 0.01 0.01 8.50 8.00 8.00
RJm 1186 1186 1188 1189 1176 1190 1200
hump bow1 level, % 62.50 63.00 47.40 65.20 65.60 64.00 63.50
(PVEK~~OW rate. lb/hr 2.6 3.7 9 .O 3.9 8.6 6.8 5.6
Voids, % 0.53 0.53 0.53 0.53 0.53 0.53 0.53
Row rate ofgas, std litershin 3.30 Ne 3.30 He 3.30 He 3.30 We 3.30 He 3.30 He 3.30 He
Sample line purge Off Off Off Qff Off Off Qff
Fission product is~topeSa
Wa,f-lif~ Fission
BSOtope
(days)
yieId
i”/o)
Sr-89
5 2 .00 5.46 6.82E10 6.9 1ElO 6.75E10 5.96E10 8.04E 10 4.80E 10 4.79E10
0.770 0.747 0.730 0.645 0.731 0.625 0.532
Y-9 1 58.80 5.57 8.08E10 9.54Ei0 9.87E10 1.89Eli 7.57E10 9.42ElO 8.48~10
0.996 1.128 1.165 2.229 0.760 1.292 1.017
%a-140 12.80 5.40 1.54Ell 1.58EI 1 1.49E11 1.57Ell 1.71Ell 4.82EI0 8.80E 10
1.116 1.097 1.035 1.098 1.043 1.071 0.912
CS-1 37 10,958.00 6.58 4.01E9 4.97E9 5.06E9 4.67E9 5.94E9 4.51E9 4.8689
0.689 0.848 0.863 0.797 0.983 0.749 0.793
Ce-141 33.00 7.09 1.11Ell i.12EIi 1.O5Ell 1.02Ell 1.32E11 6.12E10 8.04E 10
0.902 0.868 0.814 0.791 0.846 0.643 0.693
C‘e-144 284.00 4.6 1 6.70E10 6.5OElO 6.55E10 5.93El0 6.82E10 6.19E10 6.03E10
1.216 1.165 1.174 1.063 1.156 1.128 1.058
Nd-147 11.10 I .98 7.16E10 6.78E10 7.52E10 6.36E10 7.28B10
1.341 1.224 1.360 1.161 1.170
Zr-95 65.00 6.05 8.31E10 a.42~10 8.33E10 8.11E10 7.89E10 7.26B10 7.64EI0
0.954 0.930 0.920 0.897 0.744 0.919 0.848
m-95 35 .00 6.05 6.4UE9 3.55E9 -2.37B9 -9.50B8 2.30E9 1.49E10 4.36E9
-0.092 0.050 -0.034 -0.013 0.030 0.186 0.054
MU-99 2.79 4.80 1 .ow1 1 1.26Ell 3.llE9 2.98E10 2.33F3 4.71E10 1.14811
0.626 0.754 0.019 0.186 0.014 3.340 0.781
Ru-I03 39.60 1.99 5.8489 9.48E9 1.25B8 l.llE9 7.9887 3.4889 3.50E9
0.180 0.279 0.004 0.033 0.002 0.289 0.1 io
Ru-106 367.00 0.43 3.24E8 6.05E8 1.14E7 8.37E7
7.81E8 2.90E8
0.058 0.107 0.002 0.015
0.140 0.05 1
Ag-l 11 7.50 0.02 2.9687
5.0187 I.16E8
0.041
0.478 0.234
Te-129111 34.00 0.33 7.24B8 1.76B9
1.38E8
9.44E8 1.49E9
0.128 0.295
0.023
0.222 0.278
Te-132 3.25 4.40 6.40E9 1.22E10 2.28E38 1.0QE9 4.67E7 6.40E9 2.64E10
0.042 0.080 0.002 0.007 0.000 0.547 0.206
1-131 8.05 2.90 3.84E10 4.78E10 5.65El0 1.29E10 3.85E10 9.49E9 2.37E10
0.444 0.539 0.640 0.148 0.402 0.648 0.407
S ~ Constituentsh P
Constituent
U-233 7.60 7.14 6.84 6.68 7.28 7.13 4.67
i.137 1.068 1.023 0.999 1.089 1.067 0.998
U-tQtal 6.84 7.41 5.78 5.94 6.38 7.44 7.32
0.848 0.918 0.716 0.736 0.791 0.922 0.907
Li 1232.0 138.9 99.3 107.1 108.4 100.6 87.7
10.669 1.203 0.860 0.927 0.939 0.871 5.759
Be . 68.2 67.5 64.7 65.8 62.4 68.7 65.3
1.021 1.010 0.969 0.985 0.934 1.028 0.978
ZP 133.5 160.5 127.7 131.0 75.8
i 10.7 170.5
1.154 1.381 1.104 1.132 0.655 0.957 1.474

Sample number 19-1 -FVS
Sample weight, g 11.2623
Date 8-11-69
Megawatt-hours 20,309.0
Power, MW 0.00
Rpm 1189
hump bowl level, % 71.50
Ove~flow rate, Ib/hr 0.0
Voids, % 0.00
Flow rate of gas. std litersimin 3.30 He
Sample line purge On
lsotops
Sr-89
Y-9 1
Ba-140
cs-137
Ce-141
Ce-144
Nd-147
Zr-95
Nb-95
Ma-99
Ru-103
WU-106
Ag-111
Te-l29m
Te-132
1-1 3 1
Constituent
61-2 3 3
U-total
Li
Be
Half-life Fission
(days)
yield
(%l-
52.00 5.46 4.2288
0.809
58.80 5.57 1.9488
0.004
12.80 5.40 1.12E7
0.004
10,958.00 6.58 1.35E38
0.025
33.00 4.09 6.92B7
0.002
284.00 4.61 4.94B8
0.010
11.10 1.98
65.00 6.05 1.60E8
0.003
35.00 6.05 8.79E9
0.111
2.79 4.80
39.60 1.99 9.53B8
0.071
369.00 0.43 2.23E8
Q.023
7.50 0.02
34.00 0.33 l.97E8
0.104
3.35 4.40
8.05 2.90 4.70E7
0.281
8.0520
0.008
0.0766
0.009
118.1
1.023
90.7
1.358
5.79
0.050
35
Table 6.8 (continued)
196-FVS 14-9-FVS 19-24-FVS 19-36-FVS 1942-FVS 1944-FVS 1947-FVS
0.2130 10.2040 1.7933
8-1569 8-1869 9-1069
20,309.0 20,309.0 22,255.0
0.00 0.01 0.01
1190 1189 1165
62.00 65.00 62.10
0.7 2.4 4.7
0.00 0.60 0.70
3.30 He 3.30 He 2.90 Ar
On On 0 ff
7.80E8
0.017
7.6988
0.016
5.35E7
0.022
5.36E8
0.100
5.49B9
0.152
1.07E9
0.021
7.02E6
0.014
5.44E8
0.010
-8.33E7
-0.001
3.88E7
0.003
1.02E7
0.002
2.08E7
0.012
0,1454
0.022
0.1454
0.018
2.46
8.021
110.80
0.958
Fission product isotope6
3.13El0
0.700
1.81E10
0.393
1.79E9
0.848
2.74E9
0.516
2.04E10
0.600
4.51E10
0.913
7.50E7
0.184
3.75E10
0.723
2.60ElO
0.343
8.48B6
0 .OO 1
5.05E10
0.886
7.58E10
1.366
1.22E10
0.190
4.67E9
0.848
4.78E10
0.748
5.OSElO
1.000
3.00E9
0.119
5.24E10
0.859
2 2 5 El 0
0.33Q
1.85E10
0.220
3.83E9
0.209
1.24E9
0.233
2.21E9
0.742
1.83ElO
0.236
2.55E110
0.596
Sait constituent&
6.54 7.06
0.978 1.056
4.13 5.26
0.512 0.652
77.4 95.4
0.690 0.826
52.1 71.6
0.780 1.042
64.40 131.1
6.559 1.133
14 .O 175
9-29-69
23,838.0
5 .50
608
58.20
0.9
0 .00
3.35 He
Off
5.24EIO
0.820
6.04E10
0.998
7.10E10
0.938
6.19E9
1.101
5.95810
0.964
5.49L10
1.068
2.71 E10
0.944
5.70E10
0.858
-5.76E9
-0.086
1.51E9
0.019
4.72E7
0.002
4.23E7
0.115
i.21E7
0.003
4.36E.8
0.006
2.54E10
0.576
6.28
0.940
6.42
0.796
86.6
0.750
61.2
0.916
115.7
1 .000
2.1662
10-3-69
24,39 1 .0
7.00
1176
68.00
6.3
0.53
3.30 He
Off
5.87EIO
0.859
7.19EiO
1.123
8S8El0
0.975
7.67E9
1.355
7.26E10
0.841
5.84P10
1.123
4.08E 10
1.214
6.531310
0.932
-4.00E9
-0.060
4.78E10
0.431
3.14E9
0.133
4.80B8
0.089
9.05E7
0.202
7.24E8
8.181
9.59E9
0.098
3.47E10
0.652
9.28
1.089
3.73
0.462
93.3
0.808
70.5
1.055
149.9
1.286
4.9789
10-669
24,949.0
8.00
1188
64.00
4.4
0.53
3.30 He
Off
5.73El0
0.780
8.24E10
1.208
1.02Ell
0.990
4.64E8
0.081
7.8SEi0
0.819
5.78E10
1 .095
4.64E10
1.174
7.52E10
1.011
3.54E4
-0.053
5.85El0
0.412
1.16E9
0.045
7.15E4
0.013
3.23B8
0.596
2.26E9
0.018
5.72E 10
0.89 1
6.94
1.038
4.93
0.611
97.3
0.842
68.4
1.024
157.6
1.362
2.1144
10-7-69
25,190.0
8.00
1175
61.40
1.8
0.53
3.35 He
Off
6.74E10
0.889
8.72El0
1.244
1.07E1 1
0.982
4.55E9
0.795
8.60E10
0.860
6.09E IO
1.117
5 .om10
1.202
7.55ElO
0.990
1.77E9
0.026
1.26Ei 1
0.829
4.1 OE9
0.153
2.92E8
0.053
5.88E4
0.102
8.99E8
0.194
8.52E9
0.064
7.36E9
0.’108
7.24
1.083
5.24
0.649
111.3
0.964
61.3
0.918
129.2
l.i 17

Sample number 17-2-FVS
Sample weight, g 53.3293
Date 1-1469
Megawatt-hours 13.5
Power, MW 0.45
Rpm 1180
Pump bowl level, 7% 59.20
Overflow rate, Ibjhr 3.8
Voids, 70 0.60
Flow rate of gas, std litersjmin 3.30 He
Sample line purge On
Half-life
Fission
yield
lsotope
SI-89
(days)
52.00
t%)
~.
5.46 1.3889
0.687
Sr-90 10,264.00 5.86 2.87E9
0.709
Y-9 1 58.80 5.57 2.43B9
0.657
Ba-140 12.80 5.40 1.3788
1.421
CS-137 10,958.00 6.58 3.00E9
0.741
Ce-141 33.00 7.09
Ce-144
Nd-147
Zr-95
“5
RIO-99
Ru-103
Ru-106
Ag-1 11
Te-129m
Te-132
1-131
Constituent
U-233
U-total
Li
Be
284.00 4.61 4.23E10
1.096
11.10 1.98
65.00 6.05
35.00 6.05
2.79 4.80
39.60 1.99
367.00 0.43
7.50 0.02
34.00 0.33
3.25 4.40
8.05 2.90
4.28E9
0.790
4.18~9
0.5 19
S.IlE8
2.481
2.1367
0.050
1.4687
0.003
1.15E6
2.573
3.69E6
0.257
5.99E8
3.675
8.78~7
1.672
6.39
0.956
7.4 I
0.918
100.9
0.874
62.9
0.942
98.90
0.855
36
Table 6.8 (continued)
11-7-FVS 17-10-FVS 17-22-FVS 17-29-FVS 17-31-FVS 17-32-NFV
6.0976
1-2369
752.0
4.40
1180
57.20
1.8
0.60
3.30 He
On
5 .5 8E9
0.477
4.59E8
0.112
1.06E10
0.964
2.64E10
0.805
2.4969
a.680
i.02~1a
0.531
2.75E10
0.702
8.87E9
0.657
9.47E9
0.717
2.34E9
0.289
9.0969
0.1 17
1.59EX
0.034
7.99E6
0.002
2.01E7
0.091
1.76E8
0.208
2.06E10
0.307
1.02E10
0.403
4.72
0.706
90.4
0.783
13.21 73 38.3200
1-2869 2-28-69
1244.0 5757.0
4.70 7.00
1180 942
56.80 64.50
1.6 0.1
0.60 0.00
3.30 He 3.30 He
On On
Pision product isotope#
1.06E10
0.596
3.12E9
0.755
1.97E10
1.224
3.33E10
0.688
3.45B9
0.833
2.83EI0
0.443
2.46E10
0.618
1.93E10
0.985
1.62E10
0.890
-2.0013
0.023
4.60E10
0.525
X.34E7
0.01 1
2.69E6
0.001
2.66E7
0.020
2.35E9
0.030
1.30E10
0.369
3.52E10
0.554
7.26E10
1.310
4.76ElO
0.384
4.37E8
0.09 8
8.34El0
0.802
4.89E10
1.072
4.65E10
0.985
5.10ElO
0.887
-1.19E9
-0.050
1.34E4
0.010
1.98E7
0.001
I .40E6
0.000
2.6386
0.004
l.lOE7
0.002
2.55E9
0.020
3.37E10
0.456
Salt constituent@
6.116 5.98
0.915 0.895
5.570 3.19
0.690 0.395
117.68 91.0
1.019 0.788
61.78 57.5
0.925 0.861
102.14 92.7
0.883 0.801
46.2666
3-26-69
9158.0
6.90
1050
57.80
1.3
0.05
3.30 He
On
6.56610
0.765
9.35B10
1.221
1.38E11
1.078
4.30E9
0.91 1
1.12EIl
0.855
6.33E10
1.276
5.45810
1.216
7.71E10
0.983
2.15E10
0.513
4.63~8
0.004
7.44E8
0.022
6.0967
0.010
1.15E9
0.012
1.98E10
0.284
7.01
1.049
4.88
0.605
110.2
0.954
61.47
0.920
110.9
0.959
57.3671
4-169
15,168 .0
7.20
1050
61.50
3 .0
0.05
3.30 He
On
5.86E10
0.630
1.20E10
2.526
5.22EI0
0.646
1.46Ell
1.050
3.89E9
0.812
I.0PEII
0.568
5.64E10
i.108
5 .60E 10
1.083
7.84B1.10
0.921
1.49E10
0.322
1.8969
0.014
6.46EI
0.002
8.59E7
0.133
1.25EB
0.019
4.78E9
0.038
2.37E10
0.305
7.88
1.179
5.916
0.733
107.0
0.926
71.23
1.066
15.25
0.650
5.6200
4-369
10,456.0
7.20
1050
60.40
4.5
0.05
3.30 He
On
7.21El0
0.758
7.9 1 E 10
0.942
1.44Ell
1.014
3.69E9
0.766
1.16E11
0.800
5.97 IS 10
1.164
6.11E10
1.157
8.37EIO
0.962
1.48E I0
0.311
4.92E8
0.003
8.28E8
0.022
5.3584
0.081
4.52E8
0.003
3.22E10
0.404
7.892
0.978
i12.1
0.97 i
63.7
0.954
134.7
1.164
“Each entry for the fission product isotopes consists of two numbers. The first number is the radioactivity of the isotope in the
sample expressed in disintegrations per minute per gram of salt. The second number is the ratio of the isotope to the amount
caIcuhted for I g of inventory salt at time of sampling.
bEach entry for the salt constituents consists of two numbers. The first number is the amount of the constituent in the sample
expressed in milligrams per gram of salt. The second number is the ratio to the amount calculated for 1 g of inventory salt at time of
sampling.

..........

Material
CGB#l I. Liquid
G

V

Nuclide Sr-89
Fission yield, % 5.86
Half-life, days 52
--
§ample
NO.
Date
19-68 l(r27-69
18-26 5-1 9-69
19-67 10-24-69
19-66 10-24-69
bration Materid
19-68 10 min Graphite
Metal
18-26 1 hr Graphite
Metal
19-67 3 hr Graphite
Metal
19-66 IOhr Graphite
Netal
Table 7.2. Deposition of fkbn products ow graphite and metal specimens
in float-window capsule immersed for various periods in MSRE pump bowl liquid At
All specimens exposed at reactor powr of 8 MW
Ba-140 Nd-147 Ce-I41 CE-144 21-95 Nb-35 M10-99 Ru-103 Ru-106 &-I11 Te-132 Te-129 1-131
5.4 1.98 7.09 4.61 6.05 6.05 4.8 1.99 0.24 0.024 4.4 0.33 2.9
12.8 11.1 33 284 65 35 2.79 39.6 367 7.5 3.25 34 8.05
__
lnvena0ry
Lkintegxafions per minute per milligram of EPI~
1.0688 1.60E8 6.0987 1.5088 5.72E7 I.02E8 9.5087 1.65E8 3.9487 3.26E6 7.8885 1.51E.8 1.51E7 9.44E7
1.2888 1.62EX 5.99E7 3.8388 S.9OE7 1.20E8 7.9487 1.43E8 4.85E7 3.3986 7.34E5 1.3288 1.85E7 R.84E1
i.02~8 1.5688 5.95~7 1.44t8 5.7687 9 . ~ 7 7.36~7 3.64E8 5.79~7 3.2386 7.77~5 i.5m 1.45~7 9.29~7
I.02E8 1.5688 5.93E7 1.4488 5.75E7 9.RSE7 7.34E7 1.6488 3.47E7 3.2386 7.76E5 1.50E8 1.4487 5.2E8
Disintegrations per minute p ~od~ed by MSRE in I kr per square sentimeter of MSRE surface
1.62ER 6.1H 2.6t8 3.IE8 2.3F7 1.388 1.288' 2.589 7.2E7 9.485 4.686 1.9F9 3.OE7 5.2E8
Deposit activity
Expressed as ratio of activity per sqriare centimeter to amount calculated For 1 mg of inventory salt at time of sampling
0.12 0.005
0.12 0.005
0.07 0.003
O.I€ 0.019
0.09 0.004
0.07 0.006
0.06 0.004
0.08 0.030
0.03 9.006
0.03 0.015
0.02 0.003
0.03 0.002
0.10
u.17
0.001
0 09 0.001
0.10 0.021
0.06 0.002
0.09 0.003
0.10 0.0004
0.98 0.004
0.08 0.001 1
0.13 0.002
0.13 0.016
0.10 0.004
~1.001 0.005
0.002 0.013
0.011 0.013
0.015 0.024
0.005 0.013
0.005 0.009
0.004

PHOTO 97t67
41
4. The noble metals run appreciably higher -- two
orders of magnitude 01 so - than the other
elements.
Salt samples taken at about the same times were
generally relatively depleted in noble metals. though
they contained amounts which appeared to vary simi-
larly from sample to sample. Thus it appears that the
surfaces were capable of preferentially removing some
nobre-metal-bearing material borne by the salt moving
through the sample shield.
5. The noble metals increased somewhat with time,
but considerabljr less than proportionateiy, as if an
initial rapid uptake were followed by a much
slower rise.
Most importantly, the increase in activity level with
time implies that the activity came from salt exposure
rather than any explanation related to passage through
the pump bowl gas, coupled with the further assump-
tion that the window improperly and inexplicably was
open.
6. As mentioned earlier, the activity ratios on the
inventory basis are about equal for the O3 >' ' 6 ~ u
and ' 32 3129mTe isotope pairs. But the longerlived
isotope is generaily somewhat lower. This is
consistent with the deposition of material whish
has been held up in the system for periods that are
appreciable but not as long as the inventory
accumulation period.
Thus foi noble metals the data of Table 7.2 imply the
accumulation from salt of colloidal noblemetal mate-
rial, which has been in the system for an appreciable
period but is carried by the salt in amounts below
inventory, but with the surface retaining much more (in
proportion to imventory) of the noblemetal nuclides
than it does the salt-seeking nuclides. For the periods of
exposure used, deposit intensities of given nuclides on
metal and graphite did not appear to differ appreciably.
7.4 Mist
One factor to be considered in the interpretation of
pump bowl sample data is the presence of mist in the
gas space' within the mist shield in the pump bowl.
There is much evidence for this, none clearer than Fig.
0 0.25 0.50 7.3, which is a photograph of a '/z -in. strip of stainless
u
BNCHES
~ig. 7.2. sample holder for short-term deposition test (fits
capsule with windows).
into O M ~ ~ H
steel (holding electron microscope sample screens) that
was exposed in the sampler cage for 12 hr. The lower
end of the 441. strip was at the salt surface.
clearly, the photograph shows that larger droplets
accumulated at lower levels, doubtless as a consequence

PHOTO 1855-74
SALT SURFACE
Fig. 7.3. Salt dropkets on il metal strip exp~sed in MSRE pump
bowl gas space for IO hr.
of a greater mist density nearer the bottom of the gas
space, at least the larger drops representing the accumu-
lation of numerous mist particles.
The mist must have been generated in one of two
ways. The first is outside the sampler shield by the
vigorous spray into the pump bowl liquid, which also
generated spray by the rising of large and small
entrained bubbles. This would be followed by drifting
42
of some of this spray mist around the spiral 1/8-in.-wide
aperture of the shield.
The second way in which mist could develop within
the sampler shield is by the rise of entrained bubbles
too fine to resist the undertow of the pump bowl
liquid. As salt entered the bottom of the sampler shield,
these bubbles would rise within the more quiescent
liquid and would generate a fine mist as they broke the
surface. ~n partacuIar,2 the liquid rushing to fill the
bubble space would create a “jet” droplet (as well as
corona droplets) which might be impelled to a consider-
able height - in the case of champagne, tickling the
nose several inches away.
The jet droplets can accumulate or concentrate
surface contaminants3’4 on the droplet surface, being
refeired to as a surface microtome by MacEntyre. Jet
droplets are likely to be about 10% of bubble diameter,
thereby a few tens of microns in diameter. It is not
certain how much of the mist within the sampler shield
was produced from outside and how much from within;
however, at least some of the mist must have been
produced within the shield, and this explanation ap-
pears sufficient to account for the phenomena ob-
served.
Kohn2 shows that fine graphite dust was carried from
the surface of molten kiF-BeF2 (66-34 mole %> by jet
droplets, with the implication that nonwetted colloidal
material on the fuel surface could similarly become
gas-borne ~
Thus a plausible mechanism for the transport of
noblemetal fission products by mists could involve the
transport of unwetted colloidal material in the salt. This
could be transported to the surface of the liquid within
the sampler shield, by rising bubbles, and should
accumulate there to some extent if surface outflow
around the spiral was impeded. A jet droplet would
concentrate this material significantly on its surface,
and any receiver in the gas phase would indicate such
concentration. Because most droplets most of the time
would settle back into the surface, the accumulation
of noble-metal-bearing colloidal material on the surfice
would not strongly be altered by this, and accumulation
would continue until by various mechanisms inflow and
outflow quantities became balanced.
References
1. J. Ha. Engel, P. W. Maubenreich, and A. Houtzeel.
Spray, Mist, Bubbles, and Foam in the Molten Salt

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18-21
19-25
18-29
18-42
19-13
19-14
19-15
19-16
19-19
19-20
19-23
19-28
19-29
19-37
19-38
19-41
19-46
19-54
19-56
19-62
19-64
19-65
19-70
19-73
19-77
19-78
19-79
20-9
20-12
20-21
20-32

Fuel salt was co?ltl~UQtISly sprayed through the pimp
bowl purge gas to permit reniovril of all possible xenon
and krypton fission poducts. The purge gas passed into
the off-f-gas hies and thence lo ckarcoal beds.
The examination1 of surfaces exposed in the gaseous
rcgion above the liquid within the sampler shield spiral
in the pump bowl indicated the presence of appreciable
concentrations of noble metals. raising a question as to
what might actually be in the pump howl gas. (The data
or1 surfaces exposed in the pump bowl are in a separate
section.)
8.1 Freeze-Valve Capsule
48
8. GAS SAMPLES
The transfer tube and spray shield region above the
pump bowl liquid level were not designed as a facility
for sanipling the pump bowl gas. However, the inven-
tion of a freeze-valve capsule sampling device2 (men-
tioned earlier for salt samples) made it possible to
obtain useful samples from the gas region within the
spray shield. It was required that the sampling device be
small enough to pass freely through the bends of the
1'I2-in. diam. saiiipling pipe and that it should operate
automatically when it reached the pump bowl.
The device, the original form of which is shown in
Fig. 8.1, operatcd satisfactorily to furnish 20-cc samples
of gas. The capsule was evacuated and heated to 600"C,
then cooled under vacuum to allow the EiZ BeF4 in the
seal to freeze. The double seal picvented loss of
Liz BeF4 from the capsule during sampling. The
weighed, evacuated capsule was lowered into the pump
bowl thiough the salt sampling pipe and positioned
with the bottom of the capsule 1 in. above the fuel salt
level for 10 min. The freeze seal melted at the 600°C
pump bowl temperature, and gas filled the 20-cc
volume. The capsule was then withdrawn to 2 ft above
the pump bowl and allowed to cool.
The cooled resealed capsule was withdrawn from the
sampling pipe and trarisported in a carrier to the
analytical hot cells. A Teflon plug was placed over the
protruding capillary, and the exterior of the capsule was
thoroughly leached free of fission product activities.
The top of the capsule was cut off, and the interior
metal surface was leached with basic and acid solutio~ls
for 1 hr. The bottom of the Capsule was then cut at two
levels to expose the bottom chamber of the capsule.
The four capsule pieces were placed in a beaker and
thoroughly leached with 8 N IIN03 until the remaining
activity was less than 0.1% of the origina! activity. The
three leach solutions were analyLed radiochemically.
VOLUME
20cc-
ORNL-OWG 69-4784A
-- STAINLESS STEEL
CABLE
Fig. 8.1. Freeze valve capsule-
3/,-in. OD NICKEL
NlCKEL CAPILLARY
Using this device the first gas sample was obtained in
late December 1966, during run 10. In all, eight such
samples were taken during operation with U fuel.
Data for these sanrples are shown in Table 8.2 (at the
end of this chapter).
8.2 Validity of Gas Samples
There are at least two particular questions that should
be addresscd to the data obtained on gas samples.
First, do the data indicate that a valid sample of
pump bowl gas was obtained? Second, what fraction of
the MSRE production of any @veri fission product
chain is represented by flow to off-gas of a gas of the
indicated composition?

.. .
'*w.p'
The first question was approached by estimating the
activity of 50-day *'Sr resulting from the stripping of
3.2-min "Kr into the pump bowl purge gas. For
example. the activity of full-power samples 11-46,
11-53, and 13-26 averaged 3.8 X IO9 dis/niin fm 89Sr,
indicating that the gas contained 2.0 X IOL3 atoms of
"Sr per cubic centimeter of capsule volume. If the
only losses of "Kr were by decay or 100% efficient
stripping in the pump bow!, then
49
Two other gas samples (1 1-42 and 12-7) were taken
while the system had been at negligible power for several
hours. Concentrations of Sr were about an order of
magnitude iower (0.06 and 0.23) in accord with the
above viewpoint; of course. purely gas samples should
go to zero, and purely salt samples should &e unaffected
within such a period.
Thus we conclude that the samples are quite possibly
valid gas samples, probably with some mist involve-
ment.
The salt-seeking elements. including zirconium cer-
ium, and uranium, do not involve volatilization as a
means of entering the gas phase. It is sliown in Table
8.2 that an individual gas sample contains quantities of
these elements equivalent to a common magnitude of
inventory salt. Thus the conjectured presence of salt
mist in the gas samples appears verified.
The noble metals appeared in the gas samples in
quantities that are orders of magnitude higher (in
proportion to inventory salt) than was found for the
salt-seeking elements.
Thus appreciable quantities of noble metals are
involved in the gas-phase samples. If they are rapidly
stripped as a result of volatility, then the observed
concentrations represent the stripping of a major part
of the fission products in this way. However, the
volatile compounds of these elements thermo-
dynamically are not stable in the fbel salt. If the noble
metals are not volatile, then some form of aerosol mist
or spray is indicated. The relationship in this case
between aerosol concentrations within the sampler
shield, in the gas space above the violently agitated
pump bowl liquid surfacet and in the shielded
approaches to the off-gas exit from the pump bowl have
not been established. IBowever,3 in the pump bowl
proper, "the spray produced a mist of salt droplets,
some of which drifted into the off-gas line at a rate of a
few grams per month" (3.6 g/nionth is equal to ICY8 g
of salt per cubic centimeter of off-gas flow). As we skd1
see, in the samples taken during U operation, the
quantity of gas-borne salt mist in our samples, though
low, was higher than this.
"Kr stripped/min -
--,
F
' Kr produced/min F + X
where X is the decay corrstant of 3.2-min 89Kr
(0.2177 niin -I), F is the fraction of fuel volume
that passes through the pump bowl per minute;
for 50 gpm spray flow and 15 gpni fountain
flow and a salt volume of 72 cu ft, F 4.121
Thereby, F/'i(F+ X) = 0.357. At nominal full power of 8
MW. and a fission yield of 4.79%: the rate of production
of "Kr is 7.25 X BO1 atoms/min. Thus 2.59 X 10"
atoms/min enter the pump bowl. Helium purge flow at
the pump bowl temperature and pressure is 8370
cc/min, so that the gases mix for a concentration of
3.09 X 10' atoms of *'Kr per cubic centimeter. With
'agas space-in the pump bowl of 54,400 cc, the average
concentration in the well-mixed pump bowl is 1.28 X
atoms of *' Kr per cubic centimeter. ?lie
difference between this and the entrant concentration
represents the "R'n and *9Sr produced in the pump
bowl gas phase. Doubtless most of these atoms return
to the salt.
The observed concentration, 2.0 x loe3 atonis of
" ~ r per cubic centimeter of capsule volume, is
somewhat greater than the calculated average " Kr
concentration in the pump bowl gas (maximum, 1.3 X
atorn~of~~I(r per cubic centimeter of pump bowl
gas). Possibly some of the "Rb and "Sr atoms from
"Kr decay in the pump bowl remained gas-borne,
entering with the sample. The concentration of "Sr
in the fuel salt was about 1.2 X 1016 atondg.
Thus all of the observed 89Sr activity in these
gas samples corresponds to about 2 mg of fuel
salt per cubic centimeter of gas. Howevert the 23567
contents of these samples were 9, 23,and 25 pg 8.3 Double-Wall Freeze Valve Capsuk
per 20-cc sample, and the salt contained about 15,088 A number of gas samples were taken during
17
pg of 2 3 5 ~ per gram, implying entrained fuel salt operation using the freeze-valve capsule, with capsule
amounts of 0.03, 0.08, and 0.88 mg/cc. Very possibly, volume of 30 cc; results are shown in Table 8.3. These
mists containing fuel salt could remain stable within extended across run 19. However, many aggressive acid
the relatively tranquil sampler shield for a time suffi- leaches of the capsule were required to reduce external
cient to accumulate on their surfaces much of the gasborne
" RI~ and ' Sr and to enter the capsule with the
activities to values assuredly below the contained
sample, and sometimes leakage resulted. To relieve this
sample gas.
and also to provide a more certain fusible vacuum seal,

the double-walled capsule sampler shown in Fig. 8.2
was employed. This device contained an evacuated
copper vial with an 1 internal nozzle sealed by a soldered
ball. The nozzle tip was inserted through and welded at
the end to an outer capsule tip. A cap was welded to
the top of the outer capsule. completeIy protecting the
inner capsule from Contamination. In practice, after a
sanaple obtained lasing this device was transferred to the
High Radiation Level Analytical Laboratory, the tip was
abraded to free the nozzle, and the upper cap was cut
off, permitting the inner capsule to siide directly out
into a clean container for dissolution without touching
any contaminated objects. Usually the part of the
nozzle projecting from the internal capsule was cut off
and analpzed separately.
Samples 19-77, 19-79. 20-9, 20-12,20-27, and 28-32,
in addition, employed capsules in which a cap con-
taining a metal felt filter (capable of retaining 100% of
4-p particles) covered the nozzle tip. This served to
reduce the amounts of the larger mist particles carried
into the capsule.
Data from all the gas samples taken during the 33 U
operation are shown in TabIe 8.3. Reactor operating
conditions which might affect samples are also shown.
The data of Table 8.3 show the activity of the various
nuclides in the entire capsule (including nozzle for
71
P\
CUT FOR
SAMPLE REMOVAL 7 f NICKEL OUTER TUBE
ABRADE FOR
SAMPLE REMOVAL
BALL RETAINER
ORNL-UWG 70-6758
ALL WITH SOFT SOLDER SEAL
COPPER NOZZLE TUBE
Fig. 8.2. Double-wraBI sample capsule.
doubie-walled capsules), divided by the capsule volume.
Some attributes of a number of the samples are of
interest.
Samples 19-23, 19-37, and 19-56 were taken after the
power had been lowered for several hours.
Samples 19-79 and 20-32 were taken after reactor
shutdown and drain. Some salt constituents and noble
metals still remain reasonably strong, implying that the
salt mist is fairly persistent.
Sample 70 was taken “upside down”; strangely, it
appeared to accumulate more salt-seeking elements.
Samples 14-29, 19-64, and 19-73 were i6~~ntro19’
samples: the internal no2zle sear, normally soldered,
was instead a bored copper bar which did not open. So
data are only from the nozzle tube, as no gas could
enter the capsule.
8.4 Effect of Mist
As it was evident that dl samples tended tu have salt
mist, daughters of noble gases. and relatively hi&
proportions of noble metals in them, a variety of ways
were examined to separate these and to determine
which materials, if any, were truly gas-borne as opposed
to being components of the mist. It was concluded that
the lower part of the nozzle tube (external to the gas
capsule proper, but within the containment capsule)
would carry mostly mist-borne materials; some of these
would continue to the part of the nozzle tube that
extended into the internal capsule, and of course was
included with it when dissolved for analysis. The
amounts of salt in nozzle and capsule segments were
estimated for each sample by calculating and averaging
the amuunts of “inventory” salt indicated by the
various salt- seeking nuclides.
For all nuclides a gross value was obtained by
summing nozzle and capsule total and dividing by
capsule volume. The “net” value for a given nuclide was
obtained by subtracting from the observed capsule
value an mount of nozzle mist measured by the
salt-seeking e~ements and nuclides contained in the
capsule; the amount remaining was then divided by
capsule volume.
This was done for all gas samples taken at power
during runs E9 and 20. The results are shown in Table
8.1 expressed as fractions of MSRE production indi-
cated by the samples to have been gas-borne. with gross
values including. and net values exciuding, mist. Median
values, which do not give undue importance to occa-
sional high values, siiould represent the data best,
though means are also shown.
The median values indicate that only very slight net
amounts of noble metaIs (the table indicates O6 Ru as a
’&

.

h
07 0 e2 G a3 v,
5
9
3
g, 8
c? 9t 1
i c- c

E P
w

Sample number
Capsule volume, cc
Date
Megawatt-hours
Power, MW
Rpm
Pump bowl level, %
Overflow rate, Ib/hr
Voids, 7%
Flow rate of gas, std fiters/min
Sample line purge
Half-life Figdon yield
amtope (bys) B %I
SF-89 52.00 5.46
Y-98 58.80 5.57
&a-140 12.80 5.40
C'S-I37 10958.00 6.58
(*e-1 4 I 33.00 7.09

Sample nwmber
Capsule volume. cc
Date
hnegwat t-hours
Power, MW
b m
Pump bowl Bevel, %
Overflow rate, Ib/hr
Voids, 70
Flow rate of gas, std Biters/min
Sample line purge
Isotope
Sr-89
Y-9 1
Ba-140
cs-137
Ce-14 1
Ce-144
Nd-147
2-95
I%-95
h40-99
Ru-103
Ru-106
Ag-B 11
Sb-125
Te-129m
Te-132
1-131
Constituents
u-233
IBe
Zr
(day SB
52.00
58.80
12.80
1095 8 .OO
33.00
284.00
11.10
45.00
35.00
2.79
39.60
367.00
4.50
986.00
34.00
3.25
8.05
f%)
5.46
5.57
5.40
6.58
7.09
4.61
1098
4.05
6.05
4.80
B .99
0.43
0.02
0.08
0.33
4.40
2.90
FP19-I 3
7.80
8-21-69
203 E 0 .O
0.01
1165
63.25
0.1
0.38
3.30 He
OR
5.49E6
I28
5.63B5
12.676
7.47E5
139
2.12ES
6.631
6.47B5
13.213
1.16F.4
33.378
4.33ES
8.615
22x7
298
B BE7
105 3
3.22E6
613
1.9286
1238
1 .os E6
13462
1.62ES
1650
0.001 2
182
0.221 8
27,484
0.001 5
13.320
FP19-14
15 .00
8-21-69
203 10 .0
0.01
1165
64.10
0.1
0.38
3.30 Me
On
5.65E6
132
8.93E5
20.166
9.40B4
52.222
3.5086
652
1.37EQ
43.187
4.5 9 E6
93.605
3.09E6
61.487
4.27E7
575
6.1985
6472
1.26E7
1104
3.6OE6
688
2.57E6
1662
2.20EB
24329
2.07B5
2113
0.0004
61.041
0.001 1
132
0.0108
93.504
0.01 14
171
0.5253
4540
56
FP19-15
7.80
8-2 1-69
20310.0
0.01
1165
63.34
0.1
0.38
3.30 He
On
2.68E7
628
3.12E6
70.644
1.55B5
87.151
2.00E6
472
2.6386
51 529
3.49B6
71.167
4.21E4
B 24
2.27B6
45.385
2.42E8
3270
1.82E4
18482
3.28E7
2910
9.65E6
1845
1.05E6
26311
7.92E6
5160
4.49E6
53997
4.83E5
4957
Sdt crmrtituentsb
FP19-16
15.00
8-2 1-69
20410.0
0.01
1165
65.75
0.1
0.38
3.30 ~e
OR
7.87E6
185
4.48E5
10.159
3.35B5
I89
1.63F.6
304
3.4XE6
110
8.13E6
166
1.20E5
356
5.13B6
103
3.h8E7
429
8.4QE.5
8317
5.96E6
529
2.47E6
280
8.07E4
30 I
3.07E6
2008
3.40E6
43478
4.98E5
5108
0.0007 0.001 1
91.398 B 66
0.0010
I24
0.0064 0.0143
55.500 127
0.001 I Q.0096
16.889 144
FPIS-19
7.80
94-69
21557.0
5.50
1100
62.00
1 .5
0.50
2.40 AP
Off
2.9487
569
1.27E6
25.158
1.58B6
33.409
3.04E7
5565
1.45E5
2.739
5.22ES
10.457
3.35E5
17.799
4.64E5
8.141
1.33E7
192
1.64E8
2075
9.74E6
616
8.2IE5
155
1.46E5
2637
2.04E4
73.063
2.9BE6
1142
1 .ME8
1518
3.59E7
1088
0.0002
25.894
c.oom
953x1
0.0009
13.435
FP19-20
15 .OO
9469
2 B 587 .O
5 .so
1100
60.20
1.5
0.50
2.40 Ar
Off
2.28%
43.931
6.61E5
13.044
4.99E5
10.346
3.65B6
668
1.19E5
2.222
6.12ES
12.265
1.3985
7.222
3.43E5
6.012
1.73B6
24.916
1.16E8
1432
4.84%
304
6.67E.5
226
1.76B5
609
1.41E6
563
9.9387
1385
5.8987
1749
0.0001
13.166
0.0006
94.349
U.0018
26.946
FP19-23
15.00
9-1049
22255.0
0.01
1165
66.40
6.8
0.70
2.40 ar
Off
4.65E6
81.348
3.43E6
61.751
8.20E4
B 26
1.37E7
2480
4.00E6
46.8(12
4.59E6
90.646
1.29F6
46.615
2.09E6
34.261
8.53E6
125
1.68E8
1900
5.9587
3221
5.44E6
I022
1.99B5
537
2.29E6
966
6.3 1 E7
782
1.01E8
2314
0.0m5
91.215
0.001 3
165
0.0035
30.014
0.0025
36.926
"P19-28
15.00
9-23-69
23434.0
5.50
1188
67.00
7.5
0.53
2.40 Efe
Off
9.07E-7
B430
7.6785
12.757
3.28%
41.677
1.O1E6
180
2.16E5
2.809
2.36B5
4.600
2.lIC5
6.945
B .38E5
2.088
4.09E9
608
2.08E9
22584
6.87E7
3228
3.29E4
613
24x6
6092
1.4987
4183
6.65E8
$018
1.22E8
2498
0.0000
6.284
0.0030
25.894
0.001 2
13.964

Sample number
capsule volume, cc
Wee
Megawatt-hours
Power, MW
RPrn
Pump bawl level, lo
@erflow rate. iblhr
Voids, %
mow Faate of Bas, std liters/ ‘train
$ample line pusge
Sr-89
Y-9 f
h-140
cs-137
Ce-141
Ce-144
Kd-147
Zr-95
Nd-95
Ma-99
Ru-I03
wu-106
Ag-11 1
Te-129M
Te-132
L13f
Constibment
u-233
kl-total
Li
Be
52.00 5.46
58.80 5.57
12.80 5.40
10958.06) 6.58
33.00
284.00
11.10
65.00
35.00
2.79
39.60
367.00
7.50
34.00
3.25
8.05
7.09
4.61
1.98
6.05
4.05
4.80
1.99
0.43
0.02
0.33
4.40
2.90
FPI9-29
7.80
9-23-69
23458.0
5.50
1188
64.00
3.9
0.53
2.40 He
Off
7.23E6
114
8.71E5
14.436
1.8SE6
23.177
7.1385
128
5.29E5
6.859
9.36E5
18.244
2.8585
9.362
3.63E5
9.481
5.95E6
88.787
3.17E8
3409
1.7787
825
1.15E6
2% 3
7.5 1 E5
1841
I .05 E7
2922
5.5188
6594
1.56E8
3218
0.0002
28.388
0.0005
4.440
0.0010
15.354
FP19-34
15 .OO
9-30-69
23936.0
0.01
611
64.60
0.5
0.06)
2.45 He
Off
7.67B6
119
4.68E5
7.635
5.1985
6.745
6.03E6
1072
1.90E5
2.402
9.07E5
17.639
8.00E4
2.749
1.63E4
24.315
4.1 387
619
5.8888
7332
3.9587
1803
3.55%
660
7.2QE5
1930
1237
3436
1.47B8
2026
2.77E8
6190
0.0002
35.907
0.0008
11.976
5’7
FP19-38
15.00
161-Id?
24025.0
5.50
610
60.80
0.9
0.00
2.40 lie
Off Off
FP1941
15.00
10-369
24352.0
7 .00
1175
64.00
4.3
0.53
2.40 He
Fission product isotope@
2.40B7
369
3.65E4
0.591
6.5 8E5
8.361
3.ME5
53.996
2.9f F3
0.036
1.31E4
0.254
2.5184
0.372
4.1485
6.238
2.19E7
25 7
2.97E6
134
B.97E5
36.589
3.22B5
836
1 0lEA
242
2.11E7
355
2.3387
506
6.29B6
92.451
3.55E5
5.568
1.21E6
13.870
1.69B6
299
1.85E5
2.165
2.64E5
5.071
9.27E;B
2.791
3.44B5
4.976
3.29B7
49 I
6.93E8
6420
8.07EB
344
2.45E6
453
7.B3E5
1621
1.24E7
3128
4.46E8
4660
7.60E7
1450
0.0000 0.0001
2.693 10.074
0.0009
116
0.0017 0.0014
25.948 20.958
FPB946
15 .00
10-769
25153.0
8.00
1176
63.80
5.8
0.53
3.30 He
Off
1.51E8
2001
4.67%
66.858
1.33B7
123
2.03E6
354
3.53E.6
35.582
2.25E6
42.310
1.41E6
33.733
1.91%
25.175
2.47E7
367
9.53E.8
6356
5.7987
21 65
3.25B6
592
5.OIE6
8749
4.63E-l
10109
2.48E9
18644
1.9588
2580
0.0004
54.857
0.0039
484
0.0046,
39.458
0.001 7
25.948
PP19-54
15 .00
10-1449
26601 .O
8.00
1185
67.05
8.3
0.53
3.30 Be
Off
2.6986
30.537
1.6585
2.038
1.64B5
1.188
5.59E.5
95.991
1.19E.5
0.973
2.79E5
5.054
1.00E4
1.318
1.37B5
1.575
2.8386
40.848
2.81E7
162
5.56E6
143
l.BSE6
204
8.47E4
116
9.0785
161
3.61E6
23.015
1.2186
14.092
0.0000
6.882
0.0006
8.982
PPI9-56
15 .0Q
10-15-69
26821.0
0.01
1185
66.05
7 .o
0.53
3.35 He
Off
1 llE6
12.310
2.99E5
3.628
5.69E.5
4.067
4.3585
74.429
3.8BE5
3.045
2.47E5
4.465
2.35E5
4.346
1 .@E5
1.910
3.74E.6
53.736
l.tllE8
1063
7.13EB
21 7
4.35E5
77.01 7
102E5
139
2.77B6
483
8.53E7
554
8.93B6
1D 3
0.000
4.987
0.0028
41.916
FBI942
I! 5.00
10-22-69
28673.0
8.00
I186
63.60
3.9
0.53
3.30 He
Off
2.56E7
259
4.61ES
5.104
1.91E6
12.544
4.73&6
498
2.55B5
1.837
I.54E5
2.402
1.93E5
3.328
9.07E4
0.943
3.40%
46.894
1.51E8
928
6.87B6
188
4.08E5
70.941
1.57B5
205
3.63E6
546
1.77E8
1190
B.77E7
194
0.0000
5.187
0.0079
68.687
0.001 B
15.968

Sample number
Capsule volume, cc
Date
[alegawa tt-hours
PoweF, MW
Rpm
Pump bowl level. %
Overflow rate, b/hr
Voids, %
How rate of gs, std iiterq'min
Sample line purge
Wf'4ife Fission yield
Bmtope (days) d 70)
Ss-89 52.00 5.46
11-91 58.80 5.57
Ba-140 12.80 5.40
CS-B 39 10958.00 6.58
Ce-141 33.00 7.09
cc-144 2R4.00 4.68
Nd-147 I1.IID 1.9X
Zr95 65.0(1 6.05
Nb-95 35.00 6.05
MU-99 2.49 4.80
WU-103 39.60 1.99
~~-106 367.00 0.43
Ap-Ill 7.50 0.02
le-129111 34.00 0.33
Te-132 3.25 4.40
1-131 8.05 2.90
Constituent
U-233
&I-total
Li
Be
FPBSa-64
7.80
10-22-49
281 82.0
8.00
1188
63.00
3.4
0.53
3.30 We
OFF
4.27E6
42.821
1.92E5
2.i 86
I .388.:6
9.050
2.8286
475
2.41 1'5
1.919
I .64liS
2.874
I .901:5
3.243
2.2418.5
2.3L8
I .67 1.6
22.825
8.46B7
516
2.96E6
80.203
2.66085
34.775
3.46B5
45 1
1.2887
1984
7.65E8
5103
6.23E7
678
0.0001
10.358
0.0154
I33
0.0023
34.546
K'P19-65
15 .00
10-23-69
28299.0
x.oo
1185
63.40
4.0
0.53
3.25 He
Qff
4.86E7
481
6.80E5
7.424
4.86E6
31.558
1.5386
25 8
1.3483
0.097
E .44 E4
0.256
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61
Fig. 9.2. Surveillance specimen stringer.
Fig. 9.3. Stringer containment.
PHOTO 8167!
SWBE TUBE

within the basket along the specimens, no detailed flow
analysis has been presented, but quite likely the flow
may have been barely turbulent, because of entrance
effects which persisted or recurred along the flow
channel.
The first two times that the surveillance assemblies
were removed from the MSRE (after runs 7 and 1 1) for
fission product examinations, metal samples were ob-
tained by cutting the perforated cylindrical Nastelhy N
basket that held the assembly. The next two times
(after ruiis 14 and 181, '/'-in. tubing that had held
dosimeter wires was cut up to provide samples. When
the tubing was first used, lower values were obtained
(after run 11) for the deposition of fission products.
After run I& the basket was no longer of use and was
cut up to see if differences in deposition were due to
differences in type of sample.
The distribution of temperatures and of neutron
fluxes aiong the graphite sample assembly were esti-
mated2 for a I] operatiorr (Fig. 9.5). The temperature
of the graphite was generally 8 10 10°F greater than
that of the adjacent fuel, normally which entered the
channel at about 1180°F and emerged at 1210°F. The
thermal-neutron flux at the center was about 4.5 X
1013 with a fast flux of 11 these values
declined to about one-third or one-fourth of the peak at
the ends of the rig.
62
Fig. 9.4. Stringer assembly.
itmi
It was found on removal of the assembly after run 7
that mechanical distortion of the stringer bundle with
specimen breakage had occurred, apparently because
tolerance to thermal expansion had been reduced by
salt frozen between ends of consecutive graphite speci-
mens. The entire assembly was thereby replaced, with
slight modifications to design; this design was used
without further difficulty until the end of run 18.
After the termination of a particular period of reactor
operation, draining of fuel salt, and circulation and
draining of flush salt (except after run 18, when no
flush was used), the core access port at the top of the
reactor vessei was opened, and the cylindrical basket
containing the stringer assemblies was removed, placed
in a sealed shielded carrier, and transported to the
segmenting cell of the High Radiation Level Exaniina-
tion facilities. Here the stringer assembly was removed,
and one or more stringers were disassembled, being
replaced by a fresh stringer assembly. Graphite bars
from different regions of the stringer were marked on
one face and set aside for fission product examination.
Stringers replaced after runs 11 and 14 contained
graphite bars made from modified and experimental
grades ~f graphite in addition to that obtained directly
from MSRE core bars (type CGB). After runs 7 and 1 I,
6 -in. rings of the cylindrical 2-in. containment basket
were cut from top, middle, and bottom regions for
. ..

.. . i.;.; ..... &. -:
fission product anaIysis. Samples of the perforated
Hastelloy N rings were weighed and dissolved. A similar
approach was used after run 18.
After runs 14 and 18, seven sections of the ‘4-h.-
diam (0.820-in.-wall) HasteIloy N tube containing the
flux monitor wire, which was attached along one
stringer, were obtained, extending from the top to the
bottom of the core. These were dissolved for fission
product analysis. This tube was necessarily subjected to
about the same flow conditions as the adjacent graphite
specimens. The flow was doubtless less turbulent than
that existing on the outside of the cylindrical contain-
ment basket.
For the graphite specimens removed at the end of
runs 7, 11, 14, and 18, the bars were first sectioned
transversely with a thin carborundum saw to provide
specimens for photographic, metallographic, autoradio-
graphic, x-radiographic, and surface x-ray examination.
The remainders of the bars - 4 in. Bong for the middle
specimen, 2’4 to 3 in. for the end specimens - were
used for milling off successive surface layers for fission
product deposition studies.
A ”planer” was constructed by the Hot Cell Opera-
tions group for milling thin layers from the four long
surfaces of each of the graphite bars. The cutter and
collection system were so designed that the major part
of the graphite dust removed was collected. By cum-
63
paring the collected weights of samples with the initial
and final diaensions of the bars and their known
densities, sampling losses of 4.5%> and 9.1%
were indicated for the top, middle, and bottom bars of
the first specimen array so examined.
The pattern of‘ sampling graphite layers shown in Fig.
9.6 was designed to niinimize cross contamination
between cuts. The identifying groove was made on the
graphite face pressed against graphite from another
stringer in the bundle and not exposed to flowing salt.
After each cut the surfaces were vacuumed to minimize
cross contamination between samples. The powdered
samples were placed in capped plastic vials arid weighed.
The depth of cut mostly was obtained from sample
weights, though checked (satisfactorily) on the middle
bar by micrometer measurements. In this way it was
possible to obtain both a “pprofiile” of the activity of a
nuclide at various depths within the graphite and also,
by appropriate surnrnation, to determine the total
deposit activity related to one square centimeter of
(superficial) graphite sample surface. Tbie profiles and
the total deposit intensity values, though originating in
the same measurements, are rriost conveniently dis-
cussed separately.
9.8.2 Relative deposit intensity. En order to compare
the intensity of fission product deposition under
various circumstances, we generally have first obtained
0 (0 20 38 46 50 6Q 70 80
DISTANCE ABOVE BOTTOM OF HORIZONTAL GRAPHITE GRID (In 1
Fig. 9.5. Neudacn flux and temperature profiles foop COB^ s~~eilhrpce assembly.

--- -
62
60
58
CIA
I
64
, -IDENTIFYING GROOVE
,"
066C in - - -
--d.
Fig. 9.6. Scheme for milling graphite samples.
ORhL- BWG 66- 11 378
TGP FRAPdlTE
MIDQLE GRAPHITE
BDTTOM GRAPH1 TE

the observed activity of the nuclide in question per unit
of specimen surface (observed disintegrations per
minute per square centimeter). For the same time, rhe
MSRE inventory activity was obtained and dividied by
the total MSWE area (metal, 0.79 X lo6 cm2 ; graphite
channels, edges, ends, and lattice bars, 2.25 X lo6 em2,
for a total of 3.04 x cm2>? giving an inventory
value (disintegrations per minute per square centimeter)
that would result if the nuclide deposited evenly on all
surfaces. The ratio of observed to inventory disintegrations
per minute per square centimeter then yields a
relative deposit intensity which may be tabulated along
with the inventory disintegrations per minute per
square centimeter.
The relative deposit intensities, of course, will average
between 0 and 1.0 when summed over all areas,
indicating the average fraction of inventory deposited
on the reactor surface. The values may be compared
freely between runs, nuclides, regions, kinds of surfaces,
and qualities of flow. The fraction of the total
inventory estimated to be deposited on a particular
surface domain is the product of the relative deposit
intensity attributed to it and the fraction of the total
area it represents. In particular, all the metal of the
system represents 26% of the tutal area, graphite edges
a similar amount, and flow channels (and lattice bars)
about 48%.
Results of the examination of metal and graphite
samples from surveillance arrays removed after runs 7,
1 It 14, and 18 are shown, respectively, in Tables 9.1,
9.2, 9.3, and 9.4: expressed in each case as relative
deposit intensities.
For each sample set the metal specimens are listed
from the reactor top tu the bottom, followed by
graphite specimens. The nuclides with noble-gas precursors
are followed by the salt-seeking isotopes,
followed by noble metals and ending with tellurium and
iodine.
A number of generalizations of interest may be noted.
As a base line, in the absence of minute salt particles
which might have come with a sample, recoils from
fission in adjacent salt should give a relative deposit
intensity of 0.001 to 0.003. The salt-seeking elements
(Ce, Nb, Zr) on both metal and graphite and the
elements with noble-gas precursors (Sr, Y, Ba, Cs) on
metal specimens do show such levels of values: even
generally being higher near the core midplane, consistent
with the higher flux.
For graphite, the elements with noble-gas daughters
have some tendency to diffuse into the graphite in the
form of their short-lived noble-gas precursor. Thus the
values for 89Sr are mostly in the 0.10~-0.20 range,
6 5
indicating that appreciablc entry has indeed taken
piace. Values for 14'Ba arid 91Y are an order of
magnitude lower, consistent with noble-gas precursors
of much shorter half-life; their values are still about an
order of magnitude ,zbove those of the salt-seeking
elements. On specimens of pyrolytic graphite, which
had appreciably less internal porosity, the entry perpen-
dicular to the graphite planes was less than that where
the edges of the planes faced the salt; both directions
were lower than CGB graphite. The data for 7Cs were
more than an order of magnitude lower than for
strontium ~ though the half-lives of the noble-gas pre-
cursors were similar (3.18 niin for "Kr and 3.9 min for
137 Xe). The inventory data used in the tabulation were
for all material built up since first power operation;
such an inventory is severalfold too high fur the later
runs in cases such as this where transport rather than
salt accumulation is important. However, inventory
adjustment is not sufficient to account for the low
levels of 'Cs values. Appreciable diffusion of cesium
from the graphite, as discussed later, appears to account
for the iow values.
The noble meials (Nb, Mo, Ru, Ag, Sb, Te) as a group
exhibited deposits relatively more intense than other
groups on graphite and even more intense deposits on
metal, where the relative intensities on various samples
approached or exceeded 1 and were in practically all
cases above 0.1. Significant percentages of the noble
metals appear to have been deposited on system
surfaces.
Generally the deposit intensities on the 2-in.-OD
perforated container cylinder (after runs 7, I 1, arid 18)
were higher than on the 1/8-in.-OD flux monitor tube,
which was attached to a stringer within the container.
Flow was turbulent around the container cylinders but
was less turbulent and may have been laminar along the
internal tube.
The most intensely deposited elements appear to have
been tellurium, antimony, and silver; on the container
cylinder the relative deposit intensities were mostly
above 1. The deposit intensities for flux monitor tubes
were at least severalfold greater than for graphite, which
should have had similar flow, so that we can conclude
that these elements had definitely more tendency to
deposit on metal than on graphite under comparable
conditions.
To a degree only slightly less intense, molybdenum
exhibited similar behavior to that described above.
Niobium appeared to have deposited with roughly
similar intensities on graphite and metal; deposition
became somewhat less intense and varied proportion-

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factor related to the permanent adherence of deposited
material to a given surface, the so-called “sticlting
factor,” is included. usually followed by the assumption
that for lack of data it will be assumed that all metal
and graphite surfaces have equal kalues of unity -
whatever hits, sticks and stays.
9.2 Find Assembly
9.2.1 Design. The final surveillance specimen array,
inserted after run 18 and removed after run 20, was of a
different design’ 4 from those previously used, in order
to include capsules containing substantial quantities of
2 3 3 ~
and other isotopes to determine accurately their
neutron capture characteristics in the MSRE spectrum.
The cylindrical geometry permitted the inclusion of a
surveillance array consisting of sets of paired metal and
graphite specimens with differing axial positions, sur-
face roughness, and adjacent flow velocities. Because
flow conditions were the same or essentially so for
metal-graphite pairs, the hydrodynamically controlled
mass transport effects, if simple, should cancel in
comparisons, and differences can be attributed to
differences in what is commonly called sticking factor.
The sample pairs are discussed below in order of
increasing turbulence.
A photograph of the final surveillance specimen
assembly is shown in Fig. 9.7. The individual specimens
and the flow associated with them will be considered
next.
9.2.2 Specimens and flow. In the noncentral regions
of the core, the flow to a fuel channel had to pass
WASTELLOV Al BASKET
/
70
through the grid of lattice bars, and according
measurements reported’ y6 on models, the velocity
the channels was 0.7 fps with a Reynolds number
to
in
of
1000. However, the flow varied with the square root of
head loss, implying that nonlarninar entrance conditions
extended over much of these channels.
The lattice bars did not extend across the central
region. The flow through central fuel channels was
indicated by model studies to be 3.7 gpm, equivalent to
2.66 fps, or a Reynolds number of 3700; the associated
liead loss dine to turbulent flow can thus be calculated
as 0.45 ft. In this region were also the circular channels
for rod thimbles and surveillance specimens; the same
driving force across the 2.6- to 2.0-in. annulus yields a
velocity of 2.6 fps and a Reynolds number of 3460.
These flows are clearly turbulent.
Flow in the circular annulus around the surveillance
specimen basket essentially controlled the pressure
drops driving the more restricted flows around and
through various specimens within the basket.
At the bottom of the basket cage was a hollow
graphite cylinder (No. 7-3, Table 9.5) with a l’t’8-in~
outside diameter and a ’&-in. inside diameter, contain-
ing a ‘h-in.-OD Hastelloy N closed cylinder (No. 7-1,
Table 9.6). The velocity in the annulus was estimated as
0.27 fps, with an associated Reynolds number of
DVp/p = 0.0104 X 0.27 X 141/0.00528 = 75; this flow
was, therefore, clearly laminar. This value was obtained
by considering flow through three resistances in series -
respectively, 20 holes in parallel, in. m diameter, ‘4
in. long; then 6 holes in parallel, ‘4 in. in diameter, ‘4
Fig. 9.7. Final survei8iance specimen assembly.
PHOTO 96504

71
Table 9.5. Relative deposition intensity of fission products on graphite surveillance specimens fram final core specimen array
Observed dpm/cm2/(MSRE inventory as isotope dpm/MSRE total metal and graphite area, em2)
Activity and inventory data are as of reactor shutdown 12/12/69
Numbers in parentheses are (MMSRE inventory/MSRE total area), dpm/cm2
B, M, and T in sample numbers signify bottom, middle, and top regions of the core
Roughness Centimeters from
89sr 1 3 7 ~ ~ 14QBa 141Ce
144Ce 95z~ 95Nb 99M0 IQ3Ru lQ6Ru lZsSb 13ZTe 129ffl
Type Sample No.
Te
core center
(1.37Ell") (8.53E9a) (1.73Ell) (1.83Ell) (8.05E10') (1.35Ell') (1.14Ell) (2.26E11) (4.48E10a) (4.47E9') (5.63B8) (1.85E10) (1.06E1 1)
Outside
(transition flow)
Outside with wire
(turbulent flow)
Inside tube
(transition flow)
Postmortem: MSRE
core bar segment
5
25
125
5
25
125
5
5
125
5
5
125
-29
-27
-25
+8
+10
4-12
-28
26
-24
+9
+I1
+13
7-3-1-B outer
7-3-144 outer
7-3-1-T outer
12-1-B outer
12-1-M outer
12-1-T outer
7-3-1-B inner
7-3-1-M inner
7-3-1-T inner
12-1-B inner
12-1-M inner
12-1-T inner
4.2 0.023
1.6 0.022
2.4
1.9
2.0
1.8
0.31 0.0058
0.26 0.0016
0.18 0.0015
0.49 0.0029
0.34 0.0010
0.39 0.0032
0.17
0.13
0.08
0.17
0.17
0.16
0.016
0.009
0.009
0.046
0.028
0.033
0.0039
0.003 1
0.0069
0.0090
0.0037
0.001 2
0.0012
0.0012
0.0031
0.001 8
0.0010
0.0036
0.0008
0.0013
0.0015
0.0005
0.0006
0.0006
0.0011
0.0009
0.0009
0.0002
0.0021 0.21
0.0019 0.21
0.0018 0.18
0.0025 0.25
0.0023 0.15
0.0021 0.15
0.0014 0.04
0.0016 0.25
0.0016 0.24
0.0027 0.25
0.001 1 0.20
0.0018 0.09
0.083
0.033
0.035
0.040
0.039
0.803 0.022
0.22 0.150
0.19 0.064
99Tc
0.049
0.056
0.035
0.084
0.069
0.033
Q.035
0.0001
0.039
0.019
0.108
0.049
0.047
0.029
0.065
0.23
0.011
0.046
0.029
0.012
0.025
0.050
0.065
0.054
0.579
0.06 1
0.057
0.050
' 'Te
(Hnv = 2.9B9)
0.56 0.44
'Inventories sham accrue from all operation beginning with original startup. To correct inventories to show the material produced during current period (runs 19 and 20) only, multiply by factors given below. Ta obtain similarly corrected
deposition intensity ratios, divide the value in the table by the factor for the isotope. Factors are: 52-day 89Sr = 0.90; 59-day 91Y = 0.86; 40-day Io3Ru = 0.95; 65-day 95Za = 0.84; 284-day '44Ce = 0.36; %-year Io6Wu = 0.32; 30-year 137Cs =
0.14. For isotopes with shorter half-lives, corrections are trivial.
1311
0.0805
0.0059
0.0035
0.0033
0.0041
0.0050
0.0033
0.0026
0.0010
0.003 1
0.0019
0.0017

Type
Centimeters
Roughness from
‘pin.’ core center
5
25
125
Outside (transition flow) +24
Outside Vyith wire (turbulent flow) 5
25
125
Wire
Wire
Inside annulus (laminar flo~.)
5
125
Inside tube [transition (?) flow] 5
5
125
Stagnant (inside, liquid region)
Stagnant (inside, gas region)
Postmortem
MSRE heat exchanger segment
MSRE rod thimble segment (core)
+25
+27
+16
+18
+20
Table 9.6. Relative deposition intensity of fission products 011 IIasteMoy N surveillance specimens from final core specimen array
Observed dpm/crn2/(MSRE inventory as isotope dpm/hlSKE total metal and graphite area, tin')
Activity and inventory data are as of reactor shutdown 12/12/69
Numbers in parenthcses are (MSRE inventory/MSRE total area), dpm/cm2
B, M, and T in sample numbers signify bottom, middle, and top regions of the core
Sample
NO.
”sr 137c,
(1 37Ella) (8.53E9‘)
140Ba
(1.73Ell)
I4lCe 144c,e
(S.83Ell) (8.05E10‘)
” Zr
(1.35Ella)
95fu-b
(1.14Ell)
9 9 ~ 1~ 0 3 ~
(2.26b11) (4.48E10‘)
~ 1 0 6 ~
(4.47E9‘)
~ 1 2 5 ~
(5.63E8)
~ 13%e
(2.01Ell)
129m..
le
_-___
14-3-&
14-3-M
14-3-T
13-2-B
13-2-lLr
13-2-T
0.0016
0.0013
0.0010
0.001 3
0.0020
0.0039
0.0015
0.0056
0.0006
0.0009
0.0253
0.0022
0.001 2
0.0009
0.0007
0.0013
0.00 18
0.0030
0.0009
0.0008
0.0006
G.0009
0.001 1
0.0012
0.0004
0.0003
0.0003
0.0004
0.0005
0.0005
+18 13-3 wire 0.0019 0.0006 0.0015 0.0010 0.0001 0.0005 0.21 0.85 0.14 0.23 0.17
0.0004
0.0003
0.0003
0.0003
0.0007
0.0004
+11 12-2 wire 0.0030 0.001 1 0.0021 0.0016 0.0008 0.0008 0.88 1.7 0.25 0.19 0.79 0.09
-28
-26
+16
+16
+20
+26
+24
+29
+28
+27
7-143
7-1-’r
13-1-B
113-1-bl
I 3-1 -’r
14-2-L2
14-2-L1
14-2-GS
14 -2-G2
14-2-63
0.0018
0.0020
0.0021
0.0021
0.0019
0.0030
0.0010
0.14
0.47
0.41
0.0006
0.0007
0.0005
0.001 2
0.0007
0.0020
0.0002
0.0027
0.0022
0.0014
0.0015
0.0017
0.0020
0.0018
0.0015
0.0002
0.0002
0.0057
0.0077
0.0069
0.001 1
0.0013
0.0014
0.0013
0.0011
0.00003
0.00011
0.00001
0.00002
0.00004
0.0005
0.0006
0.0006
0.0006
0.0009:
0.00002
0.00006
0.000005
0.00001
0.00001
0.0006
0.0006
0.13
0.12
0.14
0.26
0.34
0.49
0.21
0.39
0.0007 0.37
0.41
0.0004 0.71
0.00001
0.00006
0.00003
0.00002
0.00086
0.005 1
0.025
0.0010
0.0012
0.0012
3.4
1.4
1.7
0.46
2.2
0.32
1.2
1.4
3.7
4.1
3.6
6 .O
3 .O
4.5
5.3
6.6
Y9T~
____
0.094
0.059
0.056
0.10
0.20
0.10
0.09
0.15
0.34
0.19
0.23
72
0.005
0.044
0.0012
0.0006
0.0009
0.127
0.078
0.079
0.09
0.17
0.08
0.06
0.23
0.32
0.19
0.17
0.007
0.043
0.0010
0.0006
0.0009
6.4
1.9
2.2
1.1
2.3
3.0
0.87
0.93
1.2
1.1
3.0
0.03
0.13
0.14
0.03
0.03
i*7Te
(Inv = 2.9E9)
1.4 1 .o
1.6 1 .o
aInventories shown accrue from all operation beginning with original startup. TO correct inventories to show the material produced during current period (runs 19 and 20) only, multiply by factors given below. l o obtain similarly corrected deposition
intensity ratios, divide the value in table by the factor for isotope. Factors are: 52-day 8’Sr = 0.90; 59-day 9* Y = 0.86; 40-day ‘03Ru = 0.95; 65-day 95Zr = 0.84; 284-day “4Ce = 0.36; 1-year Io6Ru = 0.32; 30-year 137Cs = 0.14. For isotopes with shorter
half-lives, corrections are trivial.
0.20
0.83
0.55
0.66
0.13
0.32
1311
(1.06Ell)
0.060
0.05 1
0.046
0.22
0.37
0.60
0.09
0.13
0.19
0.18
0.18
0.38
0.002
0.010
0.0008
0.0028
0.0005

in. long; then an annulus" in. wide, 5 in. Long. A
flow head loss of 0.057 ft, which should develop along
the outer part of the basket, was assumed.
In the annulus between the outside of the graphite
cylinder and the basket, the velocity was estimated to
be about 1.5 fps; the associated Reynolds number is
2200, and the flow was either laminar or in the
transition region. Some distance above, at the top of
the assembly, was a Hastelloy N cylinder (No. 14, TabIe
9.6; No. I. in Fig. 9.7) of similar external dimension,
which presumably experienced similar flow conditions
on the outside. This specimen was closed at the top and
had a double wal1. Inside was a bar containing electron
microscope screens. The liquid around the bar within
the cylinder was stagnant, and gas was trapped in the
upper part of the specimen.
Below this, above the midplane of the specimen cage
were located respectively graphite (Table 9.5, No. 12)
and double-walled Hastelloy N (Table 9.6, No. 13)
cylinders, with connecting '4 -in.-diam bores. Flow
through this tube is believed to have been transition or
possibly turbulent flow, though doubtless less turbulent
than around the specimen exterior. The exterior of the
1 -in.-OD cylinders was wrapped with -in. HasteIloy
N wire on '/2-in. pitch as a flow disturbance. Flow in
the annulus between the specimen exterior and the
basket was undoubtedly the most turbulent of any
affecting the set of specimens.
The data from the various specimens are presented in
Table 9.5 for graphite and Sable 9.6 for Hastelloy N in
terms of relative deposit intensity.
We saw no effect of surface roughness, which ranged
from 5 to 125 pin. rms, on either metal or graphite, so
this will not be hirther considered here.
9.2.3 Fission recoil. Becausc the specimens were
adjacent to fissioning salt in the core, some fission
products should recoil into the surface.8 We cdculate
that where the fission density equds the average for the
core, the relative impingement intensity of recoiling
fission fragments, (recoil atoms per square centimeter)/
(reactor production/surface area), ranges from 0.0036
for light fragments to 0.0027 for heavy fragments. The
ratio will be higher (around 0.005) where the fission
density is highest.
9.2.4 Salt-seeking nuclides. Relative deposition in-
tensities for salt-seeking nuclides (9 Zr, ' ' gle) are of
the order of the calculated impingement intensities or
less: for '§Zr, 0.001 to 0.0027 on graphite and 0.8003
to Q.0008 on metal; for l4'Ce? 0.0010 to 0.0890 on
graphite and 8.00Q6 to 0.0816 on metal. The deposi-
tion of 284-day '"Ce is consistent with this, on a
current basis after adjustment for prior inventory as
shown in the table footnotes.
There also appears to be some dependence on axial
location, with higher values nearer the center of the
core. Thus a11 of the salt-seeking nuclides observed on
surfaces could have arrived there by fission recoil ; the
fact that remaining deposition intensities on metal
surfaces are consistently less thm impingement densities
indicates that many atoms that impinge on the
surface may sooner or later return to the salt.
9.2.5 Nuclides with nrsble-gas precrarsors. The nuclides
with noble-gas precursors ("Sr, ' 37fs, r40Ba,
and, to a slight extent, 41 Ce) are, after formation, also
salt seekers. They are found to be deposited on metal to
about the same extent as isotopes of salt-seeking
elements, doubtless by fission recoil. However, noblegas
precursors can diffuse into graphite before decay,
providing an additional and major path into graphite. It
may be seen that values for "Sr, 'Cs? and 14'Ba for
the graphite samples are generally an order of magnitude
or more greater than for the salt-seeking elements.
It appears evident that the deposit intensity on
graphite of the isotopes with noble-gas precursors was
higher on the outside than on the inside, both of
specimen 7 and specimen 12. Flow was also more
turbulent outside than inside, and atomic mass transfer
coefficients should be higher. [Flows are not well
enough known to accurately compare the outside of the
lower specimen (No. 7-3) with the inside of the upper
graphite tube specimen (No. 12-1)].
Appreciably more 3.1-min "Mr and 3.9-min 37Xe
should enter the graphite than 16-sec 14'Xe or 2-sec
' Xe, but the ' 'Cs values are considerably lower
than for 8sSr. Only about 14% of the 13'Cs inventory
was formed during runs 19 and 20. With this correction,
however, * '6s deposition intensities still are less than
those observed for strontium. As will be discussed
later' the major part of the cesium formed in graphite
will diffuse back into the salt much more strongly than
the less volatile strontium; this presumably accounts for
the lower 'Cs intensity.
At first glance the "fast flow" values for 89Sr on
graphite appear somewhat high even though Kr entry
to graphite from salt was facilitated by the more rapid
flow. Similar intensity on dl flow-channel graphite
would account for the major part of the "Sr inventory,
while salt analysis for the period showed tkat the
salt contained about 82% of the "Sr. But the
discrepancy is not unacceptable, since most core fuel
channels had lower velocity and less turbulent, possibly
laminar flow.
On the inside of the closed tube the deposition of
39Sr and other salt-seeking daughters of noble gases was
much higher in the gas space than in the salt-filled
region. This is consistent with collection of 'Kr in the

gas space and the relative immobility of strontium
deposits on surfaces not washed by salt.
9.2.6 Noble metals: atiobhna and molybdenum.
Turning to the noble-metal fission products, we note
that "Nb deposited fairly strongly and fairly evenly on
all surfaces. The data are not inconsistent with postmortem
examination of reactor components, to be
described later. Molybdenunl (9 'Mo) deposited considerably
more strongly on nietal than OR graphite
(limited graphite data). Because the relative deposit
intensity of molybdenum on metal is similar to that of
8'Sr on graphite, which is attributed to atomic krypton
diffusion through the salt boundary layer, it may be
that molybdenum could also have been transported in
appreciable part by an atomic mechanism: and presumably
had a high sticking factor on metal (about I?).
Under similar flow conditions, the deposit intensity of
molybdenum on graphite is much less; hence the
sticking factor on graphite is doubtless much below
unity. Postmortem component examination found that
the Tc daughter also was more intensely deposited on
metal than on graphite.
The widely varying "9~o values for sait samples
taken during this period, however, imply that a significant
amount of 99Mo occurred along with other
noblemetal isotopes in pump bowl salt samples as
particulates. Since molybdenum was relatively high in
the present surveillance samples also, it may be that an
appreciable part of the deposition involved material
from this pool.
Because moiybdenum deposited more strongly than
its precursor niobium, an appreciable part of the
molybdenum found must have deposited independently
of niobium deposition, and niobium behavior may only
roughly indicate molybdenum behavior at best. This
may well be due to the relation of niobium behavior to
the redox potential of the sait, while molybdenum may
not be affected in the same way.
9.2.4 ~iuthenium. he rutlienium isotopcs, 031gu
and lo 6R~,
showed quite similar behavior as would be
expected, and did not exhibit any marked response to
flow or flux variations. The ruthenium isotopes appear
to deposit severalfold more intensely on metal than on
graphite. The correction of inventories to material
formed only during the exposure period will increase
the 'O6~u intensity ratios about tlirecfold but avi~
change the ' 03Ru intensity ratio very little. On such a
inventory basis the '03Ru deposition will then be
appreciably lower than that for O RU; this indicates
that an appreciable net time lag may occur before
deposition and argues against a dominant direct atomic
deposition mechanism for this element.
9.2.8 Tellurium. The tellurium isotopes Te (on
metals) and 'mTe (on graphite) show an appreciably
stronger (almost 40 times) relative deposit intensity on
metal than on graphite, indicative of real differences in
sticking factor. Deposit intensities of tellurium were
moderately higher in faster flow regions than in low
flow regions (2 times or more), possibly indicative of
response to mass transfer effects. Flux effects are not
significant.
Postmortem examination of MSWE components
showed 'Sb and 7Te deposition intensities which
were consistent with thisl except that the deposition
intensity of tellurium on the cure fue1-channel surfaces
was higher than we observe here on surveillance
specimens.
On balance, it appears that the sticking factor of
tellurium on metal is relativcly high. This might resuIt
from direct atomic deposition and/or deposition on
particulate material which then deposits selectively but
securely. Such strong intensity of tellurium in the
deposits could be the result of direct deposition of
tellurium, or of similar prompt deposition of precursor
antimony with retention of the tellurium daughter, or
both. The data do not tell.
9.2.9 Iodine. Iodine exhibits deposit intensities
which appear to be at least an order of magnitude lower
than tellurium, both for graphite specimens and, at
considerably higher levels for both tellurium and iodine,
for the metal specimens. The data for the metal surface
mostly vary with teliurium, suggesting that the iodine
found is a result of tellurium deposition and decay, but
with most of the iodine formed having returned to the
salt.
9.2.50 Sticking f~t~rs. In generai the data of the
final surveiilance assembly are consistent with those of
the earlier assemblies. The pairing of the metal and
graphite specimens in the final assembly permits some
conclusions about relative sticking factors that were
implied but jess certain in the earlier assemblies.
It appears evident, because the deposition intensity
differs for different isotopes under the same flow
conditions, that the sticking factor must be below unity
for may of the noble-metal isotopes: either on metal or
on graphite. The deposition intensity appears rather
generally to be higher on metals than on graphite and
could approach unity. En terms of mechanism, low
values of sticking factor could result if only part of the
area was active or if material was returned to the liquid,
either as atoms via chemical equilibrium processes, or
by pickup of deposited particulate material from the
surface.

kL..+
The values would be low if the inventory should be
distributed over a larger area than just the sum of
system metal and graphite areas. Such areas might
include the surface of bubbles or colioidd particles in
the circulating salt.
It is adequately clear. however, that under similar
flow conditions, the intensities of deposits on metal and
graphite surfaces differ significantly for most noble-
metal elements, with more intense deposits of a given
nuclide generally occurring on the metal surface of a
similarly exposed metal-graphite pair.
9.3 Prof&? h ad
9.3.1 Procedure. As described earlier in this section,
for most graphite surveillance specimens a succession of
thin cuts were made inward from each face to depths
frequently of about 50 mils (S.CIS0 in.). These cuts were
individually bottled. weighed. and analyzed for each
nuclide. The summations for the individual faces have
already been given in the earlier tables of this section. It
does not appear expedient to account here for the
individual samples (which would increase the volume of
data manyfold), since most of the information is
summarized in typical profiles shown below for samples
a $0 20 30 40 so
DISTBNCE FffOM SLAFACE (VIIS)
Fig. 9.8. Concentration profi6e for 7Cs in impregnated
CGB graphite, sample P-92.
removed after run 14. We note that improved hot-cell
milling techniques permitted recovery of 95 to 99% uf
the removed material for these samples, with little cross
contamination indicated.
The results of these procedures on two samples of
MSRE (CGB) graphite are shown in Figs. 9.8 - 9.14. The
sample labeled P58 was a CGB graphite exposed slightly
above the core midplane and was inserted after run 11.
Sample X-13 was exposed slightly below midplane and
was Inserted after run 7, being withdram and returned
after run E 1.
Data for the various nuclides are presented as semi-
logarithmic plots of activity per unit mass of graphite vs
cut depth. Although such a plot permits easy display of
the data, it does tend to obscure the general fact that
!O"
t - +--'-- .--
+ - I
I '

76
t -- - ~~
~-
l - I
Fig. 9.10. Concentration profile\ for and '"Oh in Fig. 9.11. Concentration profiles for "Sb and 95Z~ in CGB
pyrolytic gaphite.
graphite, X-13, double exposure.
-
i
L--..

\
k
n
10"
0 io 20 30 40 50
DISTANCE FROM SJRFCCE (rrils)
Fig. 9.12. Concentration profiles for lo3~u
77
C 40 20 30 4c 50
DISTANCE FROW SJRFACE (mils)
on two faces of Fig. 9.13. Concentration profiles for 106Ru5 141Ce. and
the X-13 graphite specimen, CGB, double exposure. 44Ce in CGB graphite, sample Y-9.

1
K16 1
3 '0 2 3 30 40 53
D STAIJCE FROU 5JPFACE imilsi
pig. 9.14. Concentration profiles for 1 4 ' and ~ ~ 144~e in
two snmples of CGB graphite, X-13, wide face, and P-38.
by far the greatest amount of any nucIide was to be
found wlthin a very few mils of the surface. Conse-
quently What the fission product profiles tend to
display is the behavior of the small fraction of the
deposited nuclide which penetrates beyond the first few
mils.
Additional data on samples from this set of specimens
were obtained by Guneo and co-workers, using a
technique developed by Evans. The technique offered
78
less possibility of cross contamination of samples.
According to this technique the specimen was generally
cut longitudinally and at the midplane, and a core was
driiled to the outer surface. The core was then dued to
a cold graphite coupon, which in turn was glued to a
machined steel piston. This piston, the position of
which could be measured accurately using a dial
micrometer, fitted into a holder which was moved on a
piece of emery paper. The resulting powder was Scotch
taped in place, and the total activity of various nuclides
was determined using a gamma-ray spectrometer.
3.3.2 Rmdts. Results using this technique are shown
in Figs. 9.15---9.17.
The material found on or in the graphite doubtless
emerged from the adjacent salt. Transport from salt can
occur by fission product atom recoil from adjacent fuel
salt, bp the deposition of elemental fission products
diffusing out of salt as atoms or borne by salt as
colloids. by chemical reaction of salt-soluble species
with graphite, by diffusion of gases from salt and
deposition onto graphite! and by physical transport of
salt into graphite, either by pressure permearion of
cracks, by wetting the graphite, or by sputtering
processes due to fission spikes in salt close to the
graphite surfaces.
Of these, there is no indication of reaction of
fluorides with graphite (with niobium a possible exsep-
tion), and volatile substances are not thought to be a
factor, the nobk-gas nuclides excepted. Furthermore,
the graphite did not appear to be wet by salt.
Occasionally there was an indication that salt entered
cracks in the graphite, and this couid be a factor for a
few samples.
The nuclides most certain to be found at greatest
depths should be the daughters of the noble gases.
Profiles for 'Sr and ' 40 ~a are shown in Fig. 9.9.
9.3.3 Diffusion mechanism relatiomhips. Among the
ways in which fission products might enter into and
diffuse in graphite are (1) recoil of fission fragments
from adjacent sail, (2) diffusion into graphite of noble
or other short-lived gases originating in salt which on
decay will deposit a nonmobile daughter, (3) formation
of fission products from uranium that was found at that
depth in the graphite, (4) migration of fission products
by a surface diffirsivn mechanism.
In our consideration of any mechanisms for the
migration of fission products. we will seek (1) estimates
of the amount of activity on unit area, summed over all
depths (expressed as disintegrarions per minute per
square centimeter), (2) the surface concentration
(expressed as disintegrations per minute per gram oF
graphite): and (3) the variation of concentration with

79
March

I - ,
SAMPI F LZbILC’
- GRlhClNG -’
-.pl-
0 (3 2c 30 40 50 60 70 540 320 5% 340 35C 360 373 380 390 400 410 4X 430 440 450 460
DcFTH IN GRAQ-IITE (m~ls)
Fig. 9.16. Fission product distribution in impregnated CGB (V-28) graphite specimen irradiated in MSRE cycle ending March 25,
1968.

depth - usually expressed as the depth required to
halve (or otherwise reduce) the concentration. We
should also try to relate the calculated activity to the
calculated inventory activity of fuel salt (expressed as
disintegrations per minute per gram of salt) developed
for the exposure period.
Since 25% of the fissions occurring in salt within one
range unit wili leave the salt and doubtless enter the
adjacent graphite, and if we use as applicable to salt the
range of light and heavy fragments in zirconium,
determined as 7.5 and 5.9 mg/cm2 respectively, we can
calculate the accumulated recoil activity:
inv. dpm 0.0053 for heavy
recoir c1prnlcm2 = ~ x
g salt 8.0043 for Ii&t '
I x /
where I is the ratio of local to core average neutron
flux. (1 - 2 to 4 for core center specimens. depending
on axial position.)
Only in the case of salt-seeking nuclides is recoil a
dominant factor.
The range of fksion products entering a graphite of
density 1.86 was determinedlo as 3.07 mdcm2 for
5Y and 2.5 1 mg/cm2 for "Xe; this corresponds to
16.5 and 13.5 p, so that the penetration shouId be
limited to a nominal 0.6 mi1.
The transport of fission gases in graphite has been
reported' 3' for representative CGB graphites.
The diffusion of noble gases in graphite also involves
diffusion through boundary layers of the adjacent salt.
We will express the behavior in graphite as a function of
the entering flux, and then use this as a boundary
condition for diffusion from salt.
At steady state the diffusion of a short-lived gaseous
nuclide into a plane-surfaced semi-infinite porous solid
has been shown by Evans' to be characterized by the
foIl owing :
where
Cg = atom concentration in surface gas phase,
JG = atom flux entering surcace,
E = total porosity ofgraphite,
DG = Knodsen diffusion coefficient in graphite,
IYCV) = concentration of atoms in gas phase in pores at
depth y.
82
The surface concentration of a nonmobile daughter in
graphite resulting from the decay of a diffusing short-
lived precursor is obtained from the accumulation
expression
where C2 is the daughter concentration per gram of
graphite. Integrating this across the power history of
the run, we obtain for the activity (aa 1 of the daughter
in disintegrations per minute per gram of graphite;
further,
where
a2 o/) = A (a2 )o expi-Pvi ; c 2)
JGo = gas nuclide flux into graphite at unit power,
(inv.Irun is the activity inventory accumulated by salt
during run,
Fa = fission rate at unit PQWH in given mass,
Y = fission yield.
The total accumulation for unit surface integrated
over all depths follows as
It remains to determine the flux into the graphite,JO
at unit power, by considering conditions within the salt.
The short -lived noble-gas nuclide is generated volumetrically
in the salt: the characteristic distance
(DJX)'/~ is about 0.06 cm for a 200-sec rmclicie and
about 0.013 cm for a 10-sec nuclide. The salt in a
boundary layer at 0.013 cm from the wall of a fuel
channel with a width of 1 cm will flow more slowly
than the bulk salt and will require perhaps 100 sec to
traverse the cote. For nuclides of 10 sec or less half-life
and as an approximation for longer-lived nuclides,
KedlI3 showed that such flow terms could be neglected,
so that at steady state
where QS is the volumetric generation rate in core salt,
D, is diffusion coefficient in salt, C, is the local con-
.... . ..... ...., ..
w.:*"

centration in salt, and r is the distance from the slab
channel midplane.
Boundary conditions are :
1. at midplane:
2. at either surface:
where
r=ro, Js =Jc ,
J, = flux in salt at surface;
whereby
and
where
C, = CgKc ,
K, = Ostwald solubility coefficient
Integration and satisfaction of the midplane boundary
condition yields:
C, = C cosh ( r a ) + Q/X .
The second boundary condition evaluates e:
We may now obtain
Because coth ( r Q m ) - 1 and K, 4 (foG/Ds)li2 ,
FQ Y salt vol.
(20 = I-.-.-X----.-- x Psalt .
mass core vol.
83
We now are able to write the desired expression fur
the activity of nonmobile daughters of short-lived noble
gases which diffuse into graphite.
1. Surface activity, disintegrations per minute per gram
of graphite:
1 12
salt vol. e 4 .
x- core vol. (Un)
2. Activity per grain of graphite at any depth, disinte-
grations per minute:
3. Total accumulation for unit surface, integrated over
all depths:
Table 9.7 applies these results to the specimens
removed at the end of run 14. A comparison with
observations given earlier in Table 9.3 is commented cn
later.
A third possibility of developing activity within the
graphite is from the traces of uranium found in the
graphite.
Here the relationship is:
salt vd.
As2 (U)b) = (inv.)Iun X I X -
core vol.
35 u conc. in graphite at given depth
23~conc.in 1 gofsalt >
under the assumption that the uranium was present in
the graphite at the given location for essentially all of
the run. The 3s LJ concentration in fuel salt was about
15,5QB ppm;
---
salt vol. 71 ft3
core vol. 23 ft3 -3.
The concentration at various depths of 235U in a
specimen of CGB graphite is shown in Fig. 9.113. The
surface concentration of about 70 ppm soon fdls to

0 40 20 30 40 50
DISTANCE FROM SURFACE (mils)
Pig. 9.18. Uranium235 concentration profiles in CGB and
pydytic graphite.
84
near 1 ppm. Similar data were obtained for a11
specimens.
Table 3.8 shows that for 1446e, 95Zr, Io3Ru, and
31 I: at depths greater than about 4 mils the activity
c~uld be accounted for as having been produced by the
trace of2 U which was present at that depth.
The total quantity of * ' k;' associated with graphite
surfaces was quite small. Totd deposition ranged
between 0.15 and 2.3 pg of 35U per square centimeter,
with a median value of about 0.8 pg of' 3s U per
square centimeter. This is equivalent to less than 2 g of
235U on all the system graphite surface (about 1 OR
flow-channel surfaces) out of an inventory of 45,080 g
of a U in the system.
8.3.4 Conclusions from profile data. As an overview
of the profile data the following observations appear
valid.
Let us roughly separate the depths into surface (less
than 1 mil), subsurface (about to 7 mils), moderately
deep (about 4 to 20 mils): and deep (over 20 mils).
For salt-seeking nuclides, * Ru and ' Wu and I
the moderately deep and deep regions are a result of
235U penetration and fission. wc have noted in an
earlier section that for salt-seeking nuclides the total
activity fur unit surface was in fair agreement with
45' I----
recoil effects. Profiles indicate that aImosf all of the
activity for these nuclides was very near the surface,
consistent with this. For these the only region for
which evidence is not clear is the subsurface (1 to 7
Table 9.9. Caleuhted specimen activity parameters after run 14 based on
diffusion calculations and salt inventory
-.-
~ .-
. . - . ..
Chain 89 91 137 140 141
~.
- .-
Gas half-life, sec 191 9.8 234 16 1 .I
DG. crn'isec
I),, cm2/sec
1.SE ~ 5
1.4F. ~ ~ 5
1.5E - 5
1.4E 5
1.2E - 5
1.3E - 5
1.2E - 5
1.3E - 5
1.2E 5
1.3E - 5
E 0.1 0.1 0.1 0.1 0.1
Halving depth, mjJs 56 13 55 14 4.7
Daughter "Sr SIY 1 3 7 ~ ~ 1 4 " ~ ~ "1Ce
HIiventorp/ dis/min per gam of salt 1.1El1 1.3E11 4.2Ega 1.7Ell 1.5Ell
Surface concentration,b disinrin per grain of graphite
'Total activity,b dis Inin-' cm-?
1.2E11
4.6 E I a
1.4Ell
1.2ElO
5 .Ok9
1 .YE9
2.0k11
1.9E;:IO
l.8Ell
5.6E.9
'Inventory here is total at end of run 14. Carryover from prior runs is insignificant by the end of run 14 except in the case of
'Cs, where operation prior tu end oi' rim 1 k contributes about half the inventory at the end of run 14. 3 I9 days later.
h ~ ~ a local ~ relative ~ ~ flux n of e I. ~
~~ ~

... .... .@
mils), where the values, though declining rapidly, may
be higher than explicable by these mechanisms.
The nuclides having noble-gas precursors (it.> * 'Sr,
14"Ba, I4lCe, "Y) do clearly exhibit the results of
noble-gas diffusion. The slopes and surface concentrations
are roughly as estimated. The total disintegrations
per minute per square centiineter is in accord.
In the case of
' Cs, the levels are considerably too
low in the moderately deep and deep regions, indicating
that cesium was probably somewhat mobile in the
graphite. Additional evidence on this point is presented
later.
This leaves 95Nb, 99Mo, arid possibly 132Te and
129mTe. These nuclides appear to have migrated in
graphite, and in particular there is about 16, times as
much niobium as was explicable in terms of the parent
95 Zr present or the 35 U at that depth. Such migration
might occur because the nuclides were volatile fluorides
or because they form stable carbides at this temperature
and sane surface diffusion due to metd-carbon chemi-
sorption occurred. The latter possibility, which seems
most likely, seems also to explain the traces of 35 U
found having migrated into the graphite.
9.4 Other Findings on Surveillance Specimens
In visual examination of surveillance specimens, slight
amounts of flush salt and, on occasion, dark green fuel
salt were found as small droplets and plates on the
surface of specimens, particularly on faces that had
P-77
X-13 wide
X-13 narrow
Y-9
P-58
P-92
K-1 wide
K-1 narrow
PG I
PG II
85
been in contact with other specimens. A brown
dusty-looking film was discerned on about half of the
fuel-exposed surfaces, using a 30-power microscope.
Examination by electron microscopy of a surface film
removed by pressing acetone-dampened cellulose acetate
tape against a graphite surface revealed only
graphite diffraction patterns. Later, spectrographic
analysis showed that appreciable quantities of stable
molybdenum isotopes were present on many surfaces.
Presumably these were not crystalline enough to show
electron diffraction patterns.
Thin transverse slices of five specimens were examined
by x radiography. Many of the salt-exposed
surfaces and some other surfaces appeared to have a
thin film of heavy material less than 10 mils thick.
Results of spectrographic analyses of samples from
graphite surface' cuts are shown in Table 9.9? expressed
as micrograms of element per square centimeter. Data
for zirconium, lithium, and iron are not included here
since they showed too much scatter to be usefully
interpreted.
About 15 pg of fuel salt would contain 1 pg of
beryllium. Similarly, about 13 1-16 of Hastellay N might
contain 1 pg of chromium and also about 9 pg of nickel
and perhaps 2 pg of molybdenum. Thus the spectrographic
analysis for beryllium corresponds to about
6-60 pg of fuel salt per square centimeter. The nickel
analyses correspond tu 7-9 pg of Hastelloy N per
square centimeter, and the chromium and molybdenum
(except for a high value) are in at least rough
agreement .
Table 9.8. List of miUed cuts from gaphite for which the fisshrn product content could be
approximately accounted for by the uranium present
3,10 6-10 6,lO
10 7-10 7-10
10 5-10 5-10
10 IO 10
9,lO 5-10 5-10
5-7 5-7
5-9 5-9
2 3-10 3-10 3-6,s 3 4
10 7,10
4-1 0 10
7-10
5-10
IO
7-10
5-10 $-to
5-10 1
5-9
2-10 2-10 2-16) 2-10
2-10 10
"Nominally, in mils, cut No. 1 was i;
2, $2; 3, 1; 4, 2; 5, 3; 6, 5; 7, 8; 8, LO; 9, IO; IO, 10.
'The samples are listed in order of distance fFOKl the bottom of the reactor.
3-10
7-10 9-10
5-10 3-10
10 10
7-10 7-10
49-10
1,5--10
5-9
2-10 3-10
2,8--10 5-10

The quantities of molybdenum are too high to
correspond to Hastelloy N composition and strongly
suggest that they are appreciabiy made up of fission
product molybdenum.
Between the end of run 11 and run 24, about 4400
effective full-power hours were developed, and MSRE
would contain about 140 g of stable molybdenum
isotopes formed by fission, or about 46 pg per square
centimeter of MSRE surface. The observed median
value of 9 wouM correspond to a relative deposit
intensity of 0.2, a magnitude quite comparable with the
0.1 median value reported for Mo deposit intensities
on graphite for this set of stringers.
Wliquots of two of these samples were examined mass
spectrographically for inolybdenum isotopes. Table
Graphite
Sample
NR-5W
NR-SN
P-55W
P-77W
P-77N
x-1 3w
x-1 3N
Y-9w
P-58W
P-92W
P-92X
Poc 0-w
P oca-N
Pyr I
Pyr II
MR-low
MR-1 ON
Table 9.9. SpectYogaphic analyses of graphite
specimens after 32,000 M W
Micrograms pee Square Centimeter
4.1
4.6
of Spec h en
Be MO Cr Ni
0.457
1.61
1.0
1.1
0.7
0.4
0.4
2.00
2.67
0.60
0.62
1.3
4.1
17.6
44.
4.5
7.4 0.15
9.3 1.03 4.4
8.7
11.3
0.49 5.5
9.0
10.5
0.55 6.6
86
9.18 gives the isotopic composition (stable) for natural
molybdenum, fission product molybdenum (for suitable
irradiation and cooling periods), and the samples.
This table suggests that the deposits contained comparable
parts of natural and fission product molybdenum.
Some preferential deposition of the 95 and 97
chdins seems indicated, possibly by stronger deposition
of niobium precursors.
Determinations of lithium and fluorine penetration
into MSRE graphite reported by MacBdin, Gibbons. and
co-workers' 4-1 were made by inserting samples across
a collimated beam of 2.06-MeV protons, measuring the
t9~(p,cuy)'60 intense gamma ray and neutrons from
the 'Li(p.n)'Be reaction. Graphite was appropriately
abraded tu permit determination of these salt constit-
uents at various depths, up to about 200 mils.
Data for the two specimens - Y-7, removed after run
11, and X-13, removed after run 14 - are shown in
Figs. 9.19-9.25.
These data for each specimen show, plotted logarith-
mically, a decline in concentration sf lithium, and
fluorine with depth. Possibly the simplest summary is
that, for specimen X-13 removed after run 14. the
lithium, fluorine, and, in a similar specimen, 'LT
declined in the same pattern. The lithium-to-fluorine
ratios scattered around that for fuel salt (not that for
LiF). The 23 5U content. though appearing slightly high
(it was based on a separate sample), followed the same
pattern, indicating no remarkable special concentrating
effect for this element. The data for sample X-13 might
be reproduced if by some rncclianism a slight amount of
fuel salt had migrated into the gaphite
The pattern for the earlier sample Y-7, removed after
run f I, is similar with respect to lithium. &ut the
fluorine values continue to decrease with depth, so that
below about 20 mils there is an apparent deficiency of
fluorine. Any mechanism supported by this observation
would appear to require independent migration of
fluorine and lithium.
It thus appears possible that slight amounts of fuel
salt may have migrated into the graphite, and the
presence of uranium (and resultant fission products)
Table 9. IO. Percentage isotopic compo~itlo~l of molybdenum on surveillance specimens
92 94 95 96 91 98 100
Fh'atural 15.86 9.12 15.70 16.50 9.45 23.75 9.62
Fission 0 0 22 -25 0 25-24 25 -24 27-26
Samples 4.0; 5.0 2.9: 3.8 28.6; 21.3 4.8; 4.9 30.2; 29.6 15.1; 15.2 14.1; 14.4
Redetermination 3.4: 3.3 2.1; 1.8 28.9; 30.3 3.9; 3.6 32.7; 33.5 15.9; 14.9 13.1; 12.6

200
!GO
L
6 50
w
3
5
2
f
20
87
Fig. 9.19. Lithium corncentpatican ;as B function of distance from the swface, specimen Y-7.
OWL-DWG 68-2545R
500 c- t 4 0 MEASURING rNWA9I; FROM FIRST SURFACE
G OUTWARD TOlNARD SECO
10 2 5 10 2 5 100 2 5 ~006
DEPTH (rnlld
1000
500
200
L
$ io0
hi
iz 50
- >
[il
20
a
I
-
10
5
2
T
SECOND SJPFACE
ORhL-DVIG 66-TFi40A
Fig. 9.21. Fluorine concentrations in graphite sample Y-7.
I l l 1 b-
0005 004 002 005 ar 02 05 10
eX@oSed t0 mdten fMd Sdd in the MSRE fQOH ilh? ll2Onh.
Mrasurernents were made as the sample was ground away in
DEPTH (cml
iayers progxssing from the first surface to the center (open
concenm'on as a function Of disMce
from the surface, specimen X-13.
circles) and then from the interior toward the second surface
(closed triangles). l'he distances shown are as measured from
the nearest surface exposed to molten salts.

-
4000
500
200
E 400
a Q
v
50
20
f0
5
88
(0 400
DISTANCE FROM SURFACE ( mils)
ORNL-DWG 68-!4530
Pig. 9.22. Fluorine concentration as a function of distance from the surface, specimen X-13.
A
v
-
-I
fc
500
200
fOO
50
28
BO
5
OWL-DWG 68-44534
z 2 5 10 20 50 400 200 500
RISTANCE FROM SURFACE , mils
Fig. 9.23. Comparison of lithium concentrations in samples Y-7 and X-13.
. .. %. . . .... i _-

.,.,.,.,. ....
lb
m
500
200
{ 00
20
(0
5
50
20
p 10
t
LY
2
f
89
ORNL-BWG 68-44532
4 2 5 10 20 50 COO 200 500 too0
DISTANCE FROM SURFACE (mils)
Fig. 9.24. Mass concentration ratio, F/Li, vs depth, specimen X-13.
t 2 5 to 20 50 too 208 530
DISTANCE FROM SURFKE (mils)
Pig. 9.25. Comparison of fluorine concentrations in simples
Y-7 and X-13, a smooth line having been drawn through the
data points.
within the graphite may largely be explained by this.
Microcracks, where present, would provide a likely
path, as would special clusters of graphite porosity.
References
1. W. H. Cook, “MSRE Material Surveillance Tests,”
MSR Program Semiannu. Progr. Rep. Feb. 28, 6965,
ORNL-3812, pp. 83-86.
2. W. H. Cook, “MSRE Materials Surveillance Testing,”
hfSR Progrmn Semiannu. Bogr. Rep. Aug. 31,
194.5, QRNL-3872, p. 87.
3. C. H. Gabbard, Design and Constmcfioion of Core
Irradiation-Specimen Array for MSRE Rum 49 arid 20,
ORSL-TM-2443 (Dec. 22, 1969).
4. S. S. Kirslis, F. F. Blankenship, and L. L.
Fairchild, ”Fission Product Deposition on the Fifth Set
of Graphite and Hastelloy-N Samples from the MSRE
Core,” MSR Program Semiannu. Progr. Rep. A~4g. 31,
1970, ORNL4622, pp. 68-90.
5. R. J. KedlI Fluid Dynamic Studies of the Molten-
Salt Reactor Experiment (MSRE) Core, ORNL-
TM-3229 (Nov. 19, 1970).
6. 3. R. Engel and P. N. Haubenreich. Temperatures
in the MSRE Core during SteudyState Bower Operation.
QRNL-TM-378 (Nov. 5, 19629.
7. R. H. Perry, C. H. Chilton, and S. B. Kirkpatrick,
eds., Chemical Engineers’ Handbook, 4th ed., McGraw-
Hill Book Co., Inc., New York, 1963, p. 5 -21.
8. W. C. Yee, A Study of the Ejyects of Fission
Fragment Recoils on the Oxidation of Zirconium,
OKNL-2742, Appendix C (April 1960).
9. E. L. Compere and S. S. Kirslis, “Cesium Isotope
Migration in MSRE Graphite,” MSR Program Semiannu.
Progr. Rep. Aug. 34, 1971, QRNL4728, pp.
51-54.
10. K. B. Evans III, J. L. Rutherford, and R. B.
Perez, “Recoil of Fission Products, 11. In Heterogeneous
Carbon Structures,” J. Appi. Phys. 39, 3253-67
(1968).
11. A. P. Malinauskas, J. L. Rutherford, and R. B.
Evans 111. Gas Transport in MSRE Moderator Grizphite,

1. Review of Theory and Counter Dij$iusion Experiments,
BRNL-4148 (September 1967).
12. R. B. Evans 111, J. L. Rutherford, and A. P.
Maiinauskas, Gas Transport in MSRB Moderator
Graphite, 11. Effects o f Impregnation, 111. Variation of
Flow Properties, ORWL-4389 (May 1969).
13. R. J. Ked. A Model for Cotnputi~2g the lWigratioii
of Very Short-Lived Noble Gases into MSRE Graphite,
ORNL-TM-I810 (July 1967).
14. R. L. Macklin. J. H. Gibbons. E. Ricci. T. M.
Handley, and D. R. glunec~~ “Proton Reaction Analysis
for Lithium and Fluorine in MSW Graphite,” IWSR
Program Semiannrd. fiogr~ Rep. Feb. 29, 1968, ORNL-
4254, pp. 119-27.
15. R. L. MackIin, J. B. Gibbons, and T. H. Handley,
Proton Reaction Analysis for Lithium and Fluorine in
Graphite. Using a Slit-Scanning Technique. BRNL-
TM-2238 (July 1968).
16. R. L. Macklin, J. H. Gibbons, F. I;. Blankenship.
E. Rieci. T. Ea. Handley, and D. W. Guneo, “Analysis of
MSRE Graphite Sample X-13 for Fluorine and
Lithium,” MSR Program Semiannu. Progr. Rep. Aug
31, 1968, QKNL-4344, pp. 14-50.
\.>

91
10. EXAMINATION OF OFF-GAS SYSTEM ~ ~ OR SPECIMENS M ~ O ~
REMOVED PRIOR TO FINAL SHU'FDfbW
The off-gas from the pump bowl carried some salt
mist, gaseous fission products, and oil vapors. On a few
occasions, restrictions to flow developed, and in replac-
ing the component or removing the plugging substance,
samples could be obtained which provide some insight
into the nature of the fission product burden of the
off-gas. In addition, a special set of specimens was
installed in the "jumper line" flange near the pump
bowl after run 14 and was examined after run 18.
IO. 1 Examination of Particle Trap
Removed after Run 7
The Mark I particle trap,* which replaced the filter in
off-gas line 52% just upstream of the reactor pressure
control valve, was installed in April 1966, following
plugging difficulties experienced in February and March
1964. It was repiaced by one of similar design in
September 1966. permitting its examination. The ori-
ginal plugging problems were attributed to polymeliza-
tion of oil vapors originating in the entry into the pump
bowl of a few grams of lubricating oil per day.
The particle trap accepted off-gas about 1 hr flow
downstream from the pump bowl. Figure 18.1 shows
the arrangement of materials in the trap. The incoming
stream impinged on stainless steel mesh and then passed
through coarse (1.4 p) and fine (0.1 1.1) felt metal filters.
The stream then passed through a bed of Fiberfrax and
finally out into a separate charcoal bed before continu-
ing down the off-gas lines to the main charcoal beds.
FINE METALLIC FKTER
COARSE METALLIC FILTER
LOWER WELD a. SAMPLE 97 "A'~-~\ '\, r
Black deposits were found can the Yorkmesh at the
end of the entry pipe, as seen in Fig. 10.2. The mesh
metal was heavily carburized, indicating operating
temperatures of at least 1200°F (the gas stream at this
point was much cooler). The radiation level in some
parts of this region was about B0,008 PJbr for a probe
in the inlet tube.
In addition to an undetermined mount of metal
mesh wire, a sample of the matted deposit showed 35%
weight loss on heating to 600°C (organic vapor), with a
carbon content of 9%.
Mass spectrographic analysis showed 20 wt % Ba, 15
wt 5% Sr, Q.2 wt $4 PI, and only 8.01 wt % Be and 0.05
wt % Zr, indicating that much of the deposit was
daughters of noble-gas fission products and relatively
little was entrained salt.
Gamma-ray spectrometry indicated the presence of
137Csp "Sr, Io3Ru or "'Ru, 'lomAg, "Nb, md
40 h. Lack of quantitative data precludes a detailed
consideration of mechanisms. However, much of the
deposit appears to be the polymerization products of
oil. Salt mist was in this case largely absent, and the
fission products listed above are daughters of noble
gases or are noble metals such as were found deposited
on specimens inserted in the pump bowl. One consistent
model might be the collection of the noble-metai
nuclides. on carbonaceous Inaterid (soot?) entrained in
the pump bowl in the fuel salt and discharged from
there into the purge gas; such a soot could also adsorb
the daughters of the noble gases as it existed as an
FIBERFRBX
(LONG FIBER)--,
..- A>.--.-
. -A* ,.&a-
Fig. 10.1 MSRE off-gas gutisle trap.
\
ORML-DWG 66-11444ff
/THERMOCOUPLE (TY PI
---
---

aerosol in the off-gas. Particles of appropriate size,
density, and charge could remain gas-borne but be
removed by impingement on an oily metal sponge.
Altl~ough the filtering efficiency was progressively
better as the gas proceeded through the trap, by far the
greatest activity was indicated to be in the impingement
deposit, indicating that most of the nongas activity
reaching this point was accumulated there (equivalent
to about 1000 full-power hours of reactor operation)
and that the impinging aerosol had good collecting
power for the daughters of the noble gases. The aerosol
would have to be fairly stable to reach this point
without depositing on walls, which implies certain
limits as to size and charge. Evidently the amounts of
noble metals carried must be much less than the
amounts of daughters of noble gases (barium, stron-
tium) formed after leaving the pump bowl. Barium-138
92
Fig. 10.2. Deposits in particle trap Yorkmesh.
(about 6% chain yieId: 17-Rain Xe + 32-min Cs -+ stable
Bra) comprises most of the long-lived stable barium.
Thus it appears that if the amounts of noble metals
detected by mass spectrometry were smdl enough,
relative to barium, lo be unreported, the proportions of
noble metals borne by off-gas must be small: although
real. No difficulties were experienced with the particle
trap inserted in September 1966, and it has not been
removed from the system.
10.2 Examination of OE-Gas Jumper
Line Removed after Run 14
After the shutdown of MSRE run 14, a section of
off-gas line, the jumper section of line 522, was
removed for examination.' This line, a 3-ft section of
1/2 -in.-ID open convolution flexible hose fabricated of

type 304 stainless steel with O-ring flanges on each end,
was located about 2 ft downstream from the pump
bowl. The upstream flange was a side-entering flange
which attaches to the vertical line leaving the pump
bowl, while the downstream flange was a top-entering
flange which attaches to the widened holdup line. The
jumper discussed here, the third used in the MSRE, was
installed prior to run IO, which began in December
1966.
After the shutdown, the jumper section was transferred
to the High Radiation Level Examination Laboratory
for cutup and examination. Two flexible-shaft
tools used to probe the line leading to the pump bowl
were also obtained for examination. On opening the
container an appreciable rise in hot-cell off-gas activity
was noted, most of it passing the cell filters and being
retained by the building charcoal trap. As the jumper
section was removed from the container and pIaced OR
fresh blotter paper. some dust fell from the upstream
flange. This dust was recovered, and a possibly larger
amount was obtained from the flange face using a
camel'shair brush. The powder looked like soot; it fell
but drifted somewhat in the moving air of the cell, as if
it were a heavy dust. A sample weighing approxiniately
8 nig read about 80 R/hr at "cuntact." The downstream
flange was tapped and brushed over a sheet of paper,
and similar quantities of black powder were obtained
from it. A sample wcighing approximately 15 mg read
about 200 R/hr at contact. Chemical and radiochemical
analyses of these dust samples are given later.
fie flanges each appeared to have a smooth, dullblack
film remaining on them but no other deposits of
significance (Fig. 10.3). Some unidentified bright flecks
were seen in or on the surface of the upstream flange.
Me~e the black film was gently scratched. bright metal
shuwe d through.
Shoat sections of the jumper-line hose (Fig. 10.4)
were taken from each end, examined microscopically,
and submitted for chemical and radiochemical analysis.
Except for rather thin, dull-black films, which smoothly
covered all surfaces including the convolutions, no
deposits, attack, or other effects were seen.
Each of the flexibIe probe tools was observed to be
covered with blackish, pasty. granular niaterial (Fig.
10.5). This material was identified by x ray as fuel-salt
particles. Chemical and radiochemical analyses of the
tools wit1 be presented later.
Electron microscope photographs (Fig. 10.6) taken of
the upstream dust showed rclatively solid particles of
the order of a micron or more in dimension: surrounded
by a material of lighter and different structure which
appeared to be amorphous carbon; electron diffraction
lines for graphite were not evident.
93
An upstream I-in. section of the jumper line near the
flange read 150 R/hr at contact; a sianrlar section near
the downstream flange read 350 R/hr.
10.2.1 Chemical andysis. Portions of the upstream
and downstream powders were analyzed chemically for
carbon and spectrographically for lithium, beryllium,
zirconium, and other cations. ~n addition, z35U was
determined by neutron activation analysis; this could be
converted to total uranium by using the enrichment of
the uranium in the MSRE fuel salt, which was about
33%.
Results of these analyses are shown in Table 10.1.
Analyses of the dust samples show 12 to 16% carbon,
28 to 54% fuel salt, and 4% structural metals. Based on
activity data, fission products could have amounted to
2 to 3% of the sample weight. Thereby 53 tu 22% of
the sample weight was not accounted for in these
categories or spectrographically as other metals. The
discrepancies may have resulted from the small amounts
of sample available. The sample did not lose weight
under a heat lamp and thus did not contain readily
volatile substances .
10.2.2 Radiochemical analysis. Radiochemical analy-
ses were obtained for the noble-metal isotopes ' Wg,
lo6Ru, 'oaRu, "'ha~,and'~Nb;for~~'r;fortlerare
earths 7PJd, '"Ce, and ' 4' Ce: for the daughters of
the fission gases krypton and xenon: 91 Y, "Sr, "Sr,
40Ba, and ' 7Cs; and for the tellurium isotopes
' Te, Te, and I (tellurium daughter). These
analyses were obtained on samples of dust from
upstream and downstream flanges, on the approxi-
mately 1-in. sections of flexible hose Cut near the
flanges, and on the first flexible-shaft tool used to
probe the pump off-gas exit line.
Data obtained in the examination are shown in Table
10.2, along with ratios to inventory.
It appeared reasonable to compare the dust recovered
from the upstream or downstream regions with the
deposited material on the hose in that region; this was
done for each substance by dividing the amount
deposited by the amount found in 1 g of the associated
dust.
Values for the inlet region were reasonably consistent
for all classes of nuclides, indicating that the deposits
could be regarded as deposited dust. The median value
of about 0.004 g/cm indicates that the deposits in the
inlet region corresponded to this amount of dust. A
similar argument may be made with respect to the
downstream hose and outlet dust, which appeared to be
of about the same material, with the median indicating
about 0.016 g/cm. Ratios of outlet and inlet dust values
had a median of 1.5, indicating no great difference
between the two dust samples.

94
Fig. 10.3. Deposits on jumper line flanges after r~lp 14.
PHOTO (856 - 74

95
R - 42973
Fig. 10.4. Sections OP of€-g-gat; jumper line flexible tubing and outlet tube after run 14.

Li
Be
2r
LiF
BeF
ZrF,
23SU
UF, (total)
Carbon
Fe
Cr
Ni
Mo
AI
CU
FPs (max)"
96
T~ble 10.1. Analysis ~f dust from MSRE off-gas jumper line
Inlet Flange (wt %) Outlet Flange (wt SC)
__
hs Determined Constituent As Determined Constituent
3.4
1.7
2.74
0.358
10-14
12.6
8.9
12
("2)
5.0
1.4
2
1
0.4
1
0.1
7.3
4.2
1.4
0.596
15-17
Total found 47 78
shssumes chain deposition rate constant throughout power history (includes only chains determined).
29.1
21.9
2.6
2.4
16
(4.5)
( *" 3 )
.,..... . -..- . . , ..? ..

~
Sample
Element or
Isotope
t C
1/2
Element
Li
Be
Zr
2 3 5 ~
C
Isotope
111 Ag
0 6Hu
103RU
9 ~ o
9 .SNb
952,
147Nd
144~e
14’ce
91Y
40~a
8%
137cs
”Sr
2Te
1 29mTe
1311
~
7.6 d
365 d
39.7 d
67 hr
35 d
65 d
11.1 d
285 d
33 d
58 d
12.8 d
50.5 yr
29.2 yr
28 yr
78 hr
37 d
8.05 d
Yield‘
(%)
0.0181
0.392
2.98
6.07
6.26
6.26
2.37
5.58
6.44
5.83
6.39
4.72
6.03
5.72
4.21.
0.159
2.93
Table 10.2. Relative quantities of ekments and isotopes found in ~ ff-g~ jumper line“’ ’
Inlet Dust Outlet Dust Upstream Hose Downstream Hose Flexible Tool ’ISRE lnveneory
(Per E)
(Pep g) (per ft) (per ft) (total)
Decay Rate
(10 dislrnin)
-~
0.066
0.057
0.054
0,050
23
‘“31
10
5.7
2.8
0.54
-0,013

Fig. 10.5. Deposit on flexible probe.
98
The electron microscope photographs of dust showed
a number of fragments 1 to 4 6-1 in width with sharp
edges and many small pieces 0.1 to 0.3 p in width
(IOOQ a to 3000 A).
The inventory values used in these calculations
represent accumulations over the entire power history
(tz to); the more appropriate value would of course
be for the operating period (t2 f1 ) only:
Jr2-to =IttZ tI +ltl-ro ex^ [-Ntz - till .
The second term, representing the effect of prior
accumulation, is important only when values of X(t2 -
r,) are suitably low (less than 1). SO, except for
364-day Io6Ru, 2-year '"Sb, 30-year "'Sr, and
36-year "Cs, correction is not particularly significant.
We shall frequently use the approach
ObS - QbS inv. (total)
x
inv. (present pepid) inv. (total) inv. (present period) '
where
If the prior inventory were relatively small,
or the present period relatively long with respect to
half-life, Act2 - tl ) 3 1, then the ratio
'totalli"v.(present period)
approaches 1. It can never exceed the ratio
EFPHtotd@FPH(total after prior period) .
(EFPH is accumulated effective full-power hours.)
Examination of the data in terms of mechanisms will
be done later in the section.
At the end of run 16 a restriction existed in the
off-gas line (line 522) near the pump, vvhich had
developed since the line was reamed after run 14."' To
Li
L-.,

clear the line and recover some of the material for The isotopes "Sr and ?Cs, which have noble-gas
examination, a reaming tool with a liolbw core was precursors with half-lives of 3 to 4 min, are present in
attached to flexible metal tubing. This was attached to significantly greater proportions, consistent with a
a "May pack" case and thence to a vacuum pump mode of transport other than by salt particles.
vented into the off-gas system. The bfay pack case held The "noble-metal" isotopes 95"b, "Mo, "Ru, and
several screens of varied aperture and a filter paper. The mTe were present in even greater proportions,
specialized absorbers normally a part of the May pack indicating that they were transported more vigorously
assembly were not used.
than fuel salt. Comparison with inventory is StKai&hht-
The tool satisfactorily opened the off-gas line. A small forward in the case of 2.79-day 99Mo and 34-day
amount of blackish dust was recovered on the filter
paper and from the flexible tubing.
'29mTe, since much of the inventory was formed in
runs 15 and 16. In the case of 364-day ' Rut although
Analysis of the residue on the filter paper is shown in a major part of the run 14 material remains undecayed,
Table 10.3. The total amount of each element or salt samples during runs 15 and 16 show little to be
isotope was determined and compared with the amount present in the salt; if only the "'Ru produced by
of "inventory" fuel salt that should contain or had 233U fission is taken into account, the relative sample
produced such a value.
value is high.
The constituent elements of the fuel salt appear to be The fuel processing, completed September 7, 1968,
present in quantities indicating 4 to 7 mg of fuel salt on appeared also to have removed substantially all "Nb
the filter paper, as do the isotopes 14'Ba3 '44Ce, and from the salt. Inventory is consequently taken as that
95Zr, which usually remain with the salt. It is note- produced by decay of '§Zr from run 14 after this time
worthy that 233W is in this group, indicating that it was and that produced in runs 15 and 16. Thus the
transported only as a saIt constituent and that the salt "nc~ble-metal" elements appear to be present in the
was largely from runs 15 and 16.
naaterial removed from the off-gas line in considerably
greater proportion than other materials. It wcpuld
appear that they had a mode of transport different
Table 10.3. Material recovered from M§RE off-gas from the first two groups above, though they may nut
line after K U 16 ~
Corrected to shutdown December 16, 1968
have been transported all in the same way.
There remains 8.05-day ' I. The examination after
run 14 of the jumper section of the off-gas line found
Inventory
(per milligram
of salt)
Found
appreciable I, which may have been transferred as
__-~-__l__i
'Iota1 Ratio tu inventory 30-hr ' mTe. In the present case, essentially all the
I inventory came from a short period of high power
Elements
near the end of run 15. Near-inventory values were
In milligrams
found in salt sample FP 16-4, taken just prior to the
end of run 16. Thus it appears that the value found here
indicates little ' I transferred except as salt.
Li 0.116 0.80 I
Be 0.067 0.35 5
Zr 0.116 0.47 4
U-233 0.0067 0.0396 6
Sr-89
CS-1 37
Ba-14(1
Y-9 1
Ce-144
Zr-95
Xb-35
Ma-99
Ru-106
Te-129m
1-131
Pissiam prodwts
In disintegrations per minute
2.9E4
4.1F6
4.1E6
5.2E6
4.1E7
7.4E6
9.4E6'
3.1E4
2.XE6 (9.8Ef)
2.3E4
9.8E4
3.13E8
4.31EIf
I.47Eb
1.2r.8
1.48B8
3.08E7
1.40E9
2.76E7
3.51Eb
9.2E7
1.01E6
110
110
8
23
4
4
15 o'
900
1400'
4000
10
%ventory set to zero for fuel returned from reprocessing
September 1968.
10.4 Off-Gas Line Examinations after Run 18
After run 18 the specimen holder installed after run
14 in the jumper line outlet flange was removed, and
samples were obtained.
The off-gas specimens were exposed during 5818 hr
to about 4748 hr with fuel circulation. during runs 15,
16. 17, and 18. During this period, 2542 effective
full-power hours were developed.
Near the end of run 18, a plugging of the off-gas line
(at the pump bowl) developed. Kestriction of this flow
caused diversion of off-gas through the overflow tank,
thence via line 523 to a pressure control valve assembly,
and then into the 4-in. piping of line 522. These valves
can be closed when it is desired to blow the accumu-

lated overflow salt back into the pump bowl, but are
normally open. A flow restriction also developed in line
523 near the end of run 18, and the valve assembly was
removed for examination.
Data obtained from both sets of examinations will be
described below.
The off-gas line specimen assembly was placed after
run 14 in the flange connecting the jumper line exit to
the entry pipe leading to the 4411. pipe section of line
522. The specimen holder was made of 27 in. of
'//,-in.-OD. 0.03S-inO-wall stainless steel tubing, with a
flange insert disk on the upper end. Four slotted
sections and one unslotted section occupied the bottom
17 in. of the tube; about 8 of the upper 10 in. were
contained within the '//,-in. entry pipe; dl the rest of
the specimen holder tube projected downward into the
4-in. pipe section. The specimen arrays included a
holder for electron microscope screens, a pair of
closed-end diffusion tubes, and a graphite specimen.
A hot-cell photograph of the partially segmented tube
after exposure is shown in Fig. 10.7. The electron
microscope screens were not recovered. Data from the
diffusion tubes and graphite specimen will be presented
below. In addition, two sections were cut from the
183-in. unslotted section, of particular interest because
normally all the off-gas flow passed through this tubing.
These segments of tubing were then plugged, and the
exterior was carefully cleaned and leached repeatedly
until the leach activity was quite low; the sections were
then dissolved and the activity determined. Data from
these sections are presented in Table 10.4.
From the exterior of the tube, slightly above the
uppel slots, a thin black flake of deposited material was
recovered weighing about 283 mg. The tube, after
removal of the flake, is shown in Fig. 10.8. The
underlying metal was bright and did not appear to have
interacted with the flake substance. Analysis of the
flake is show~z in Table i0.4.
The quantity per centimeter was divided by that for 1
g af flake substance for each nuciide: the general
agreement of values indicated that they were doubtless
froan the same source and that the deposit intensities on
the two sections were about 1 and 6 mg/cm respectively
(based on median ratio values). These values will be
referred to later. Observed values are also shown for
deposits on the upstream handling "bail."
Data were a h obtained for fission product deposition
on the graphite specimen (narrow and wide facesj
and on consecutive dissolved 1-in. scrubbed segments of
'I8-in.- and '&/,-in.-l[B diffusion tubes closed on the upper
ends. The upper parts of these tubes contained packed
sections of the granular absorbents AI2 O3 and NaF.
Data on these specimens are shown in Table 10.5. As
useful models for examination of the data have not
Fig. 10.7. Off-gas line specimen holder as segmented after removal, following run 18.

__ -... .-
Inventorya
per gram
of salt
Ii 113 2.89
Be 67.6 7.11
zr c
U-233 6.7 0.517
sr-89 1.332611
Y-91 1.206Ell
h-140 1.534EI 1
es-a 37 5.45OE9
CR-141 2El I
Cc-144 6.109ElO
Nd-147 -5EI 1
zr-95 1.254Ell
m-9.5 8.92OElO
Ku-l 03 4.495ElO
Ru-B 06 3.464ElO
‘I-e-1 29 1.872610
I-131 8.OElO
Table BOA. Data on samples OH segments from off-gas line specimen holder removed following run 18
---..-
Flake Tube section
.._-_-.-
1
..,_. _L__ _ _.
Tube section 2
Amount Ratio to inventory Amount per Ratio to MSRE Ratio to Amount per Ratio to MSRE Ratio to
per gram for 1 g of salt centimeter total inventory 1 g of flake centimeter total inventory 1 g of flake
0.026
0.105
0.078
1.8E12 13
8.8E9 0.07
6.4E9 0.04
2.8EIO 5
FLOE6 0.00004
5.4E7 0.0009
1.6E9 0.03
1.6E8 0.0012
1.4Ell 1.6
1.2Ell 2.7
2.5ElO 7
2.3ElO 1.2
ElWllClatS
In milligrams
CO.016 3.2E-11

Inventory for
Nuclide 1 g of salt
(dis min-l g-l)
sr-89
Y-9 1
Ba-140
CS-137
(3-144
zr-95
Nb-95
Ru-103
Ru-106
Sh-125
Te-I29
1.3E11
1.2E11
I.SEl1
5.4E9
6.1E10
I .3Et I
8.9E10
4.5ElO
3.5E9
2.9E8
1.9E10
Table 10.5. Specimens exposed in MSRE off-gas line, runs 15-18
Radioactivity on
graphite specimen ~
Radioactivity on '/&.-ID
__......
diffusion tube (dis/min) Radioactivity on 'h-in.-ID diffusion tube (disimin)
- -. .
(dis min-l cm12) Bottom
-
inch
Wide face Narrow face
Second
inch
Third
inch
Fourth
inch
A1203
granules
NaF
granules
Second
inch
Third
inch
Fourth
inch
A12Q3
granules
NaF
granules
3.6E11
l.lE9
3.5E10
4.4E9
1.2E6
5.3B6
2.8E7
I. .SEI
3.5E7
8.SB4
3.3B7
4.4E10
4.4ET
6.2E8
5.2E5
6.7ES
a
l.lE6
1.3E7
3.7E.5
3.QE8
7.9B6
3.3E7
l.lE7
2.4E7
2.5E7
1.4E8
1.5E8
1.3E9

104
Fa- 48664
Fig. 10.8. Section of off-gas Bine specimen holder showing flaked deposit, removed after K U 18. ~
been found, these data will not be considered further
here.
10.5 Examination of Valve Assembly
from Line 523 after Run 18
The overflow h e from the pump bowl opens at
about the spray baffle, descends vertically, spirals
around the suction line below the pump bowl, then
passes through the top of the toroidal overflow tank,
and terminates near the bottom of the lank. Gas may
pass through this line if no salt covers the exit into the
overflow tank or if the pressure is sufficient (due ts
clogging of regular off-gas Bines, etc.) to cause bubbling
through any salt that is present.
In addition, helium (about 0.7 std liter/min) flows in
through two bubblers to measure the liquid level. Gas
from these bubblers, along with any off-gds flow, will
pass through line 523 to enter the 4-in.-diam part of the
main off-gas line 522.
From time to time the salt accumulated in the
overflow tank was returned to the pump bowl. This was
accomplished by closing a control valve in line 523
between the overflow tank and the entry point into h e
522. The entry of the bubbler gas pressurized the
overflow tank until return of the accumulated salt to
the pump bowl permitted the gas to pass into the pump
bowl; the control valve was then reopened.
Near the end of run 18, clogging of line 523, which
had been carrying most of the off-gas flow for several
weeks, was experienced. This was attributed to plugging
in the valve assembly, and after shutdown the assembly
was replaced. The removed equipment was transported
to the High Radiation Level Examination Laboratory,
where it was segmented for examination.
The vertical 'j2-in. entry and exit lines to the valve
assembly were spaced about 39 in. apart. terminating in
O-ring flanges. Following the entry flange the line
continued upward. across, and downward to the entry
port of the automatic control valve HCV-523. Gas flow,
proceeded past the plunger into the cylinder containing
the bellows plunger seal and upward into Gshaped exit
holes in the flange, leaving the valve flange horizontally.
Curved piping then led to a manual valve (V-523).
normally open, which was entered upward. The hori-
zontal exit pipe from this valve curved upward and then
downward to the exit flange.
Examination of HCV-523 did not reveal an obstruc-
tion to flow. 41 surfaces were covered with sootlike
deposits over bright metal; samples of this were
recovered. Some samples of black dust were obtained
from sections of the line between the two valves.
111 the norndy open V-523 the exit Wow was
downward. A blob ti€ rounded black substance (about
the size of a small pea) was found covering the exit

lnventory
for 1 g
of fuel salt
Li 116
Be 67
u 6.7
(% IJ-2351
(% U-238)
SI-89
SI-90
u-9 1
Ba-140
cs-137
Ce-141
(3-144
Nd-147
Zr-95
Nb-95
Ru-103
Ru-1 06
Ag-111
Sb-125
Te-129
1-131
1.3Ell
5.4E9
1.2E1 I
I SE11
5.5E9
1.9EI 1
6.1Ei0
5.4b10
1.3E11
8.9E10
4.5E10
3.5E9
6.7E8
2.9E38
8.3B9
8.IElO
Table 10.6. Analysis of deposits from line 523 (undetermined quantity)
Expressed as equivalent grains of inventory salt
Ratio of observed value to that for 1 g of inventorv salt
___
HCV-5 23
Line
I.each Dust Dust Dust Dust Dust
0.080
1.5
0.54
12

and appreciably higher than the salt-seeking substances
(Ei, ~ e 95~r, , I4'ce, I4'ce, a47~d, OB^. etc.).
Silver-1 11 was quite high! Antimony-125 and 29m Te
were quite high. Iodine-131 was lower than r2smTe in
each sample. On bdance the pattern is not much
different than that seen for other deposits from the gas
phase, wherever obtained. The absolute quantities were
not really very large, and no analysis would have been
called for had not the plugging that developed during
the use of this line fur off-gas flow required removal.
10.6 The Estimation of Flowing Aerosol
Concentrations from Deposits on Conduit Wails
We can observe the amounts of activity or mass
deposited on off-gas line surfaces. To find out what this
can tell us about MSRE off-gas behavior, we must
consider the relevant mechanics of aerosol deposition.
We will obtain a relation between the arrioiint entering
a conduit and the amount depositing on a given
surface segment by either diffusional or thermophoretic
mechanisms. The accumulation of such deposits over
extended operating periods will then be related to
observable mass or activity in terms of inventory values,
fraction to off-gas, etc.
Diffusion coefficients of aerosol particles are calculated
using the Einstein-Stokes-~unnin~am equation.'
106
where I is the mean free path, r the particle radius, and
Q the viscosity of the gas.
The viscosity of helium6 is q = 4.23 X 10-6TTr.5/
(P.826 - 0.409).
bameter of ~
At a pressure of 5 psig and assuming a helium
collision diameter (u) of 2.2 the mean free path is
calculated from the usual formula:
1
2 = -_____
n(2)112 ?I52
where IZ is the number of atom per cubic centimeter.
values calculated for the diffusion coefficient of
particles of various dianieters in 5 psig of Iielium are
shown in Table 10.7.
10.6.1 Deposition by diffusion. The deposition of
aerosols on conduit walls frunr an isothermal gas
stream in laminar flow is the result of particle
diffusion and follows the Townsend equation,7 ,8 which
is of the form
where n is the concentration of gas-borne particles at a
distance X from the entrance and Q is the volumetric
flow rate. The coefficients aE and b, are numerical
calculated constants.
The derivative gives the amount deposited on unit
length (n,) in terms of the amount entering the conduit,
no :
Values of the coefficients are:
T&Ie 10.9. Diffusion coefficient of particles in 5 pig of helinrn
1 'i hi
1 0.819 11.488
2 0.097 70 .O1 0
3 0.032 178.91
4 0.0157 338.0
~ifiusion coefficient (cm2/seci At temperature of -
~ ~ -~
~. -
particle (A) 923°K 873°K 533°K 443°K 433°K
3
10
30
100
3 00
1,000
1.900
3.000
10,000
311,000
5.250
4.73B 1
5.26E-2
4.14E-3
5.31E 4
4.90E -5
1.74E-5
5.89B-6
4.05E- 7
1.46E- 7
4.880
4.39E -1
4.88E- 2
4.405; 3
4.93E 4
4.56E -5
1.62E-5
5.5 1 E-6
6.70E -7
I .41 E-?
2.530
2.28E- 1
2.54E-2
2.293: -3
2.58B-4
2.43B-5
8.83E-6
3.1 IE-6
4.43F. 7
1.08E-7
2.160
1.95E-1
2.17E-2
1.96E -3
2.216-4
2.09E-5
9.65E-6
2.43E- 6
4.06E-7
1.02E-4
1.920
1.73E 1
1.93E- 2
1.74F-3
1.Y7F 4
1.84E 5
6.89E -6
2.48F 6
3.81E 7
9.75B 8
'

ha&
For suitably short distance [Le,, X < 0.01 (Q/Db,)f
the exponential factors approach unity, and we may
write
B
ns =no --X 27.
e
This is valid in our case for particfes above 1000 a at
distances below about 2 m.
As a point for later reference, in the case of particles
of 1700 A in 523°K gas flowing at 3300 std an3 of
helium per minute and 5 psig,
10.6.2 Deposition by thermophoresis. Because the
gas leaving the pump bowl (at about 650°C) was
cooled considerably (temperature below 500°F in
the jumper legion). the heat loss through the con-
duit walls implies an appreciable radidl tenrpeiature
gradient. A temperature gradient in an aerosol results in
a thermally driven Brownian motion toward the cooler
region, called thermophoresis. This effect. for example,
causes deposition of sod on lamp chimney waIls. The
velocity of particles smaller than the mean free path is
independent of size or composition. At diameters
somewhat greater than the mean free path, particle
velocities are again independent of diameter but are
diminished if the thermal conductivity of the particle is
much greater than the gas. (The effect of thermal
conductivity is not appreciable in our case.)
For particles in helium the radial velocity is given as
The radial heat transfer conditions determine both
die internal radial gradient and also the axial tempera-
ture decrease for given flow conditions. Simple stepwise
models can be set up and deposition characteristics
calculated. One such model assumed -iii.-ID conduit
(l-in~-OD), slow slug flow at 3300 std cm3/min and 5
psig, with about 50 W of beta heat per liter in the gas,
an arbitray wall conductance, and horizontal tube
convective loss to a GO'C ambient. With a 650°C inlet
temperature we found the values given in Table 10.8.
We probably do not know the conditions affecting heat
loss too much better than this, which is believed to be
reasonably representative of the MSRE conditions.
It is evident that for particles of siLes above about
100 the thermophoretic effect was dominant in
107
Table 10.8. Thelrrnophoretic deposition parameters
estimated for off-gas line
____-~ ____
__I___
10 544 0.86 1.2 x 10-
50 306 (1.58 3.9 x
100 191 0.46 1.6 x
150 147 0.40 9.1 X IO4
200 132 0.36 6.8 i: LO4
producing the observed depositions (.,/no was about 6
to 16 X for thermophoresis and about 3 X
for diffusion of 1700-A particles).
The insensitivity of thermophoresis to particle size
indicates that the observed mixture of sizes could be
approximately representative of that emerging in the
pump bowl off-gas stream.
Thus the ratio i! = nJno of deposit per unit Iength.
jfS. to that entering the conduit. 110. can be estimated
on a thermophoretic basis and on a diffusional basis.
The thermophoretic effect is appreciably greater in the
regions where the gas is cooled appreciably if particle
sizes are abovz about 100 A. We will assume below that
values ofZ are available.
10.6.3 Rerationship between observed depositio~i
and reactor loss fractions. Four patterns of deposi-
tion will be described. and the lawr relating the ok-
served deposition to the reactor situation will be
indicated.
1. Stable species in constant proportion to salt
constituents follow the simple relation no = n,/Z. Thus,
given a value ofZ, the total amount entering the off-gas
system can be obtained from an observation of the
deposit on unit length. For jumper-line situations Z is
about 0.081 (within a factor of 2). Steady deposition
can be assumed.
2. A second situation relates to the transport of
nuclides which are not retained in the salt and part of
which may be transported promptly into the off-gas
system, for example, the noble metals. In developing a
relationship for this mechanism. we define 1,- to as the
total MSRE inventory of the nuclide at time t,
produced after time to. If the fraction of production
which goes to off-gas is fr(OC), it may be shown that n,
= Z X fr(OG) X Itpto. Inasrnuch as we have inventory
values at various times,

Thus
In many cases, It/Kt.- to x 1.
For many of the noble metals, ns/It e IO-' to lo-' ;
thereby
fr(OGj < - x 10-7 .
3. A third mechanism of transport is applicable to
fission products which remained dissolved in salt, with
some salt transported into the off-gas. either as discrete
mist particles or conceivably as small amounts accumu-
lated on some ultimately transported particle, possibly
as the result of momentary vaporization of salt along a
fission spike.
If we assume a continuous transport during any
interval of reactor power,
and
Is dC -PFy - xc
dt
where C is the number of atoms per gram of salt, P is
reactor power, F is the number of fissions per gram at
unit power in unit time, y is fission yield, W is the rate
of salt transport to off-gas in gram per unit time, Z =
ns/no, and E , ~
is the number of atoms in unit length of
deposit.
From these for interval i we find:
108
and
PFv
h
The equation for Cj is evidentiy simply a version of the
inventory relationship. It appears possible to carry the
second equation through the power history. If we take
into account the lack of transport when the reactor is
drained and assume that the rate of salt transported to
off-gas is constant while the salt is circulating, we may
write W j = Wfi, where fi = 0 when drained and 1 when
circulating. Then the WZ/h term can be factored out,
and the term
can be obtained by computation through the power
history in conjunction witli the inventory equation:
The evaluation of IG(t -- to), the gas deposit
inventory, has not yet been done.
4. The final case considers the daughters of noble
gases, insofar as they are produced by decomposition in
the off-gas stream. Here the mechanism changes: The
daughter atom diffuse relatively rapidly to the walls, so
that to a good approximation the rate of deposition of
daughter atonis on the walls is the rate of decay of
xenon krypton in the adjacent gas. Thus we must, for
short-lived parent atoms, define the time of flow (7)
from the conduit entrance to the point (XI in question.
Let
where p is the gas dcnsity at the point x (at T, PI, Wis
the mass inlet rate of the gas, and A is the cross-
sectional area of the conduit.
k.-.,
. .
w.. ri;" ......

\.& ;.;.$
... ,: i.....
We may now show that the accumulated activity of
the daughter in disintegrations per minute per centi-
meter at a particular point is
XI 4x1 t.- hl?(X) fr@c;)1
2(t--to)
QCC,
where fr(cSC) is the fraction of the parent production
which enters the conduit and 12(t- to) is the MSRE
inventory activity of the daughter for the period (t, to).
Note in particular that because deposition of atoms
(rather than particulates) is rapid and the daughter
atoms are formed in flow, no Z factor appears here.
Actually the daughter atoms of noble gases are
ubiquitous in that they might deposit from the gas Onto
particulate surfaces and be transported in that fashion.
A major part is contained in the salt and would of
course move with that. However, either of these
alternative mechanisms is indicated to result in much
less deposition than that which occurs by deposition
from decay of a short-lived parent in the adjacent gas.
We now proceed to examine some of the data
presented above in the light of these relationships.
10.9 Discussion of Off--Gas Line Transport
10.7.1 Salt constituents and dt-seeking nuclides.
The deposit data for salt constituent elements and
salt-seeking nuclides can be interpreted as total amounts
of salt entering the off-gas system over the period of
exposure, using the equation presented earlier, no =
iQZ.
We have available to us the deposition per unit length
on segments from the jumper-line corrugated tubing
after run 14 and the specimen holder tube after run 18.
The deposits presumably occurred by simple diffusion
of particulates to the walls, or by thermophoresis. The
therrnophoresis mechanism is over 1 OQ-fold more rapid
here. For the regions under consideration, calculations
indicate that within a factor of about 2, the ther-
rnophoretic deposition per centimeter would be about
0.001 times that entering the tube, relatively independ-
ently of particle size. A similar rate could occur by
diffusion alone for particles of IO0 A but electron
microscope photographs showed particles 1000 to 3000
A, as well as larger.
However, the only way for the off-gas to have cooled
to the measured levels at the jumper line requires radial
temperature gradients to lose the heat, and the ther-
mophoretic effect of such gradients is well established.
Consequently, we conclude the thermophoretic effect
must have controlled deposition in the off-gas line near
the pump bowl. Thus we use a value ns/tz0 = 0.001.
hnounts of &as-borne salt estimated to have entered
the off-gas system are shown in Table 10.9.
The deposit data for the samples within a given period
are passably consistent. Using the thermophoretic depo-
sition factor, the data indicate that the amounts of salt
entering the off-gas system during the given periods
were of the order of only a few grams.
The calculation for the radioactive nuclides was
crude; only the accumulation across runs 13 -14 and
17--1& was considered, and this was treated as a single
interval at average power, prior inventory being ne-
glected. These assumptions should not introduce naajor
error, however. We conclude this indicates that major
quantities of particulate material did not pass beyond
the jumper line into the off-gas system.
About 60% of the particulate material leaving the
pump bowl should be transported to the walls by
thermophoresis in the first 2 m or so from the pump
bowl.
Diffusion alone would result in rates several hundred-
fold slower (for 1700-W particles; this varies approxi-
mately inversely with the square of particle diameter).
Consequently, if this mechanism controlled the ob-
served deposits. it would imply that considerably more
particulate material left the pump bowl and passed
through the jumper line.
10.7.2 Daughters of noble gases. The deposit activity
observed for the daughters of noble gases can be used to
calculate the fraction of production of the parent noble
gas which enters the off-gas system. Diffusion of
daughter atoms is more rapid than thermophoresis, and
deposition is assumed to occur at the same tube
positions that the parent noble gas undergoes decay in
the adjacent gas.
Table 10.9. Grams of salt estimated to enter off-gas system
Runs 10-14,9112 hr
Basis of Upstream Downstream
calculation hose hose
Li 2.3
Be 0.9
Zr 0.6
U-235 1.4
Zr-9s 0.15
(:e-141 0.2
Ce-144 0.6
PJ~-147 5
Runs 15-18> 4748 hr
Specimen Specimen
holder holder
tube 1 tube 2
4.2 0.15 0.17
1.5 0.12 0.12
0.6
4
4 0.2 0.4
0.8
6
43

Table 10.10. Estimated percentage of noble-gas nuclides entering the off-gas based on deposited
daughter activity and ratio to theoretical value for full stripping
- ~.
Runs 10 -14 Runs I S19
Upstream Downstream Specimen lnoEder Specimen holder
Gas Daughter how hose tube 1 tube 2
.._I_
Percent Ratio Percent
__
Ratio Percent Ratio Percent Ratio
191-sec Kr-89 SI-89 0.8 0.06 4.0 0.28 4 .0 0.28 5.2 0.34
33-sec Kr-90 SI-90 0.04 0.04 0.17 0.19
9.8-sec Kr-91 Y-9 1 0.07 1 .OQ 0.003 0.04 0.0Q4 0.06 0.004 0.06
16-sec Xe-140 Ba-140 0.004 0.03 0.019 0.1 2 0.005 0.03 0.008 0.05
234-sec Xe-137 CS-137 1.3 0.07 6.0 0.32 4.6 0.25 5.4 0.29
The fraction of noble gas (I) entering the off-gas
system is calculated from daughter (2) activity:
fr(1) =
obs dis min-1 cm-1) inventory total
inventory total (
flow rate 1
~ a,
area A,
The values shown below were calculated assuming a
flow rate of about 80 cm3/sec and a delay ( T~> of about
2 sec between the pump bowl and the deposition point.
MI daughter nuclides which result from fie decay ofa
noble-gas isotope are assumed to remain where deposited.
The indicated percentage of production entering the
off-gas was compared with the amount calculated to
enter the off-gas if full stripping of all of the noble-gas
burden of the salt entering the pump bowl occurred,
with no entrainment in the setui~a flow, no holdup in
gmphite or elsewhere, etc. The results are indicated in
Table 10.10. The magnitudes appear plausibfe.
Most of the values for the longer-lived gases (89Kg
and 7Xe> are 25 to 32% of the theoretical maximum,
indicating that the net stripping was only partially
complete, possibly attributable to bubble return, incomplete
mass transfer, and graphite hoIdup.
The ratio values far 9oKr, "Kr, and 140Xe mostly
are 0.06 to 0.044; the net stripping appears to be
somewhat less for these shorter-lived nuclides. Slow
mass transfer from salt to gas phases could well account
for both goups.
10.7.3 MsMe metds. The activity of deposits of
noble-metd nuclides can be used to estimate the
fraction of production that entered the off-gas system.
The relationship employed is
__
observed activity
fr(off-gas) =
MSRE inventory
MSRE inventory 1
X-- XZ'
MSW period inventory
where Z is again the ratio of the amount deposited per
centimeter to the amount entering the off-gas system
(here, of the nuclide in question). This factor as before
is approximately 0.001 if the thermcaphoresis mecha-
nism is dominant.
The period of operation was long (runs 10 to I8
extended over 388 days, runs 15 to 18 over 242 days)
with respect to the half-life of most noble-metal
nuclides, so that the ratio of the MSRE inventory to the
MSWE period inventory is not much above unity (in the
greatest case, 367-day O6 Ru. it is less than 1.1 for runs
10 to 14, and for runs 15 to I8 the ratio is below 2.7,
in spite of the shorter period, the longer prior period,
and changes in fission yield).
Table 10.1 1. Estimated fraction of noblemetal
production entering off-gas system
Runs 10-14
Nuclide Upstream r)uwnstream
hose hose
Wb-95 0.000002
Mo-39 0.00010
Ru-103 0'00012
Ru-106 0.00074
Ag-1 B 1 0.00009
Te-129 0.00018
Te-132 0.00004
1-131 0.00003
Runs 15-18
Specimen Specimen
imer holder
tube 1 tube 2
0.00001 0.00003
0.00160
0.00130 0.00002
0.00450
0.0 0 2 6 0
0.000005
0.00120
0.000 29
0.00001 0.00002
0.00010 0.00430 0.00002

.
.. -.&& . . . . (..
.
111
Thus, to a useful approximation, References
obs activity per cm 1
fr(off-gas) = X
MSRE inventory Z
We tabulate in Table 10.1 1 this fraction, using for Z the
value of 0.001 as before.
These values, of course, indicate that only negligibie
amounts of noble metals entered the off-gas system.
tenths to hundredths of one percent of production. We
believe that the assumptions invoIved in the above
estimate are acceptable and consequently that the
estimates do indicate the true magnitude of’ noblemetal
transport into off-gas.
Obviously, the estimated values depend directly on
the inverse of the deposition factor, Z. If only the
diifusiori mechanism were active (which we doubt), the
estimated amounts transported into the off-gas would
be increased seveial hundredfold (300 X ?). Even in this
situation the estimated fractions of noble metals trans-
ported into the off-gas would mostly be of the order of
a few percent OH less.
We consequently believe that the observed activities
of noble metals in off-gas line deposits indicate that
only negligible, or at riiost minor. quantities of these
substances were transported into the off-gas system.
1. A. N. Smith, “Off-Gas Filter Assembly,” MrSw
Program Seemia~intl. Progo Rep. Aug. 31, 1966, ORNL-
4034, pp. 74-77.
2. E. L. Compere, “Examination of MSWE Off-Gas
Jumper Line,” MSR Program Semiannu. Prog. Rep.
A u 31, ~ ~ 1968, OWL-4344, pp. 206 -10.
3. E. L. Compere, “bxamination of Material Recovered
from MSKE Off-Cas Line,” MSR Program
Semiannu. Pa.ogr. Rep. F’eb. 28, 1965, ORNL-4396, pp.
144 45.
4. W. W. Parkinson (OHPNL Health Physics Division),
“Analysis of Black Plug from Valve in MSK6,” letter to
E. L. Compere (ORNL Chemical Tschnology Division),
Jan. 21, 1370.
5. Cited by J. W. Thomas, The Diffusion Battery
Method fiw Aerosol Pmicle Size Deter’mination,
ORNL-1649 (1953): p. 48.
6. C. F. Bonilla, Nuclear Engineering Handbook, ed.
H. Etherington, McGraw-Hill. New Vork, 1959, p.
9-25.
7. C. N. Davies, ed., AwosoZ Science, Academic, Yew
York, 1966.
8. N. A. Fuchs, The Mechanics of Aerosols, Macrnillan,
New York, 1964.

Operation of the Molten Salt Reactor Experiment was
terminated on December 12, 1969, the salt drained, and
the system placed in standby condition. In January
1971 a number of segments were removed from
selected components in the reactor system for exami-
nation. These included a graphite bar and control rod
thimble from the center of the core, tubing and a
segment of the shell from the heat exchanger, and the
sampler-enricher mist shield and cage from the pump
bowl. The examination of these items is discussed
below. It was expedient to extract parts of the original
reports in preparing this summary.
B 1 .I Examination of Deposits from the Mist Shield
in the MSRE Fuel Pump Bowl
In January 1971 the sampler cage and mist shield
were excised from the MSRE fuel pump bowl by using
a rotated cutting wheel to trepan the pump bowl top.'
nile sample transfer tube was cut off just above the
iatch stop plug penetrating the pump bowl top; the
adjacent approximately 3-ft segment of tube was
inadvertently dropped to the bottom of the reactor cell
and could not be recovered. The final ligament attach-
ing the mist shield spiral to the pump bowl top was
severed with a chisel. The assembly was transported to
the High kidiation Level Examination Laboratory for
cutup and examination.
Removal of the assembly disclosed the copper bodies
of two sample capsules that had been dropped in 1967
and 1968 lying on the bottom of the pump bowl. Also
on the bottom of the bowl. in and around the sampler
area, was a considerable amount of fairly coarse
granular, porous black particles (largely black flakes
about 2 to 5 rnm wide and up to 1 mm thick). Contact
of the heated quartz light source in the pump bowl with
this material resulted in smoke evolution and appar-
ently some softening and smoothing of the surface of
the accumulation.
A few grams of the loose particles were recovered and
transferred in a jar to the hot cells; a week later the jar
was darkened enough to prevent seeing the particles
through the glass. An additional quantity of this
material was placed loosely in the assembly shield
carrier can. Samples were submitted for analysis for
carbon and for spectrographic and radiochemical analy-
ses. The results are discussed below.
The sampler assembly as removed from the carrier can
is shown in Fig. 11 .I. All external surfaces were covered
with a dark-gray to black film, apparently 0.1 mm or
more in thickness. &.'here the metal of the mist shield
spiral at the top had been distorted by the chisel action,
black eggshell-like film had scaled off, and the bright
metal below it appeared unattacked. Mere the metal
had not been deformed, the film did not flake off.
Scraping indicated a dense, fairly hard adherent blackish
deposit.
On the cage ring a soft deposit was noted. and some
was scraped off; the underlying metal appeared unattacked.
The heat of sun lamps used for in-cell photography
caused a smoke to evolve from deposits on bottom
surfaces of the ring and shield. 'This could have been
material, picked up during handing, similar to that seen
on the bottom of the pump bowl.
At this time, samples were scraped from the top,
middle, and bottom regions of the exterior of the mist
shield, from inside the bottom, and from the ring. The
mist shield spiral was then cut loose from the pump
bowl segment, and cuts were made to lay it open using
a cutoff wheel. A view of the two parts is shown in Fig.
11 2.
In contrast to the outside, where the changes between
gas (upper half) and liquid (lower) regions. though
evident, were not pronounced. on the inside the lower
and upper regions differed markedly in the appearance
of the deposits.
In the upper region the deposits were rather similar to
those outside, though perhaps more irregular. The
region of overlap appeared to have the heaviest deposit
in the gas region, a dark film up to I mrn thick. thickest
at the top. The tendency of aerosols to deposit on
cooler surfaces (thermophoresis) is called to mind. In
the liquid region the deposits were considerably thicker
and more irregular than elsewhere, as if formed from
larger agglomerates.
Hn the a~ea of overlap in the liquid region, this kind of
deposit was not observed, the deposit resembling that
on the outside. If we recall that flow into the mist
shield was nominally upward and then outward through
the spiral, the surfaces within the mist shield are
evidently subject to smaller liquid shear forces than
those outside or in the overlap. and the liquid was
surely niore quiescent there than elsewhere. The con-
ditions permit the accumulation and deposition of
agglomerates.
The sample cage deposits also were more even in the
upper part, beconling thickest at and on the latch stop.
The deposit on the latch stop was black and hard,
between 1 and 2 mm thck. Deposits on the cage rods ". .....,... :y
y . i;'
I

113
Fig 111.I. Mist shield containing sampler cage from MSRE pump bowl.
below the surface (see Figs. 11.3 and 11.4) were quite
irregular and lumpy and in general had a brown-tan
(copper or rust) color over darker material; some
whitish material was also seen. Four of the rods were
scraped to recover samples of the deposited material.
After a gamma-radiation survey of the cage at this time,
the unscraped cage rod was cut out for metallographic
examination; another rod was also cut out for more
thorough scraping, segmenting, and possible leaching of
the surfaces.
The gamma radiation survey was conducted by
lowering the cage in 1/2-in. or smaller steps past a 0.020-
by 1.0411. horizontal colhating slit in 4 in. of lead.
Both total radiation and gamma spectra were obtained
using a sodium iodide scintillation crystal. The radiation
levels were greatest in the latch stop region at the top of
the cage and next in magnitude at the bottom ring.
Levels along the rods were irregular but were higher in
the liquid region than in the gas area even though
considerable material had been scraped from four of the
five rods in that region. In all regions the spectrum was
predominantly that of 367-day ' Ru and 2.7year
Sbt and no striking differences in the spectral shapes
were noted.
Analyses of samples recovered from various regions
inside and outside the mist shield and sampler cage are
shown in Table 11.1. The samples generally weighed
between 0.1 and 0.4 g. The radiation level nf the
samples was measured using an in-ce11 G-M probe at
about I-in. distance and at the same distance with the
sample surrounded by a '/8-in- copper shield (to absorb
the 3.5-MeV beta of the 30-sec daughter of
'

Fig. 11.2. Interior of mist shield. Right part of right segment overlapped left part of segment on left.

. ... . . . . . . ....
.:.3j>y . .
115
Pig. 11.3. Sample cage and mist shield.

pig. 11.4. ~epsits on sampler cage. Ring already scraped.
116
' O6 Ru). Activities measured in this way ranged from 4
R/hr (2 R/hr shielded) to 180 R/hr (SO R/hr shielded),
the latter being on a 0.4-g sample of the deposit om the
latch stop at the top of the sample cage.
Spectrographic and chemical analyses are available on
three samples: (1) the black lumpy materid picked up
from the pump bowl bottom, 42) the deposit on the
latch stop at the top of the sample cage, and (3)
materid scraped-from the inside of the mist shield in
the liquid region. The material recovered from the
pump bowl bottom contained 7% carbon, 31%
Hastelloy N metals, 3.4% Jk (18% BeF2), and 6% Li
(22% LiF). Quite possibly this included some cutting
debris. The carbon doubtless was a tar or soot resulting
from thermal and radiolytic decomposition of lubri-
cating oil leaking into the pump bowl. lit is believed that
this material was jarred loose from upper parts of the
pump bowl or the sample transfer tube during the chisel
work to detach the mist shield.
The hard deposit on the latch stop contained 28%
carbon, 2.8% Be (1 1% BeF,), 2.8% lii (10% LiF)? and
12% metals in approximate Hastelloy N proportions,
again possibly to some extent cutting debris.
The sample taken from the inner liquid region of the
mist shield contained 2.5% Be (13% BeF,), 3.0% Li
(11% LiF)* and 18% metals (with somewhat more
chromium and iron than Wastelloy N); a carbon andysis
was not obtained.
In each case, about 0.5 to 1% Zr (about 1 to 2%
ZrF4) was found, a level lower &an fuel salt in
proportion to the lithium and beryllium. Uranium
analyses were not obtainable; so we cannot clearly say
whether the salt is fuel salt OK flush salt. Since fission
product data suggest that the deposits built up OWF
appreciable periods, we presume that it is fuel salt.
In all cases the dominant Bastelloy N constituent,
nickel, was the major metallic ingredient of the deposit.
Only in the deposit from the mist shield inside the
liquid region did die proportions of Ni, Ms, Cr, and Fe
depart appreciably from the metal proper. "1 this
deposit a relative excess of chromium and iron was
found, which would mot be attributable to incidental
metal debris from cutting operations. Ht is also possible
that the various Hastelloy N elements were all subject
to mas transport by salt during operation: and that
little of that found resulted from cutup operation.
We now come to consideration of fission product
isotope data. nese data are shown in Table i1.2 for
deposits scraped from a number of regions. The activity
per gram of sample is shown as a fraction of MSRE
inventory activity per gram of MSRE fuel salt to
eliminate the effects of yield and power history; u
. . . . . . . .
-2s,

Sample
Radiation level
Weight Percent 3’r
Location (;Ee;;;;.) (mg) c (dis min-‘g-l)
Table 11.1. Claemicall and spectrographic analysis of deposits from mist shield in the MSWE pump bowl
Percent Li Percent Be Percent Zr” Percent NiQ Percent Mo’” Percent CP Percent Fea Percent Mna
Pump bowl 1W5) 402 3.1 El0 6.00 3.42 0.5-1.0 20. 30 2-4 l--2 0.5-1.0 0.Sl.Q
bottom 25(12b 108 1.1
Latch stop 130(60) 291 1.85 Eli 2.15 2.01 0.5-1.0 S-10 2-- 4 0.5-1.0 0.5-1.0 CO.5
180(80 365 28 w
Inside. 40(11) 179 4.7 El0 3.03 2.52 0.5-1.0 5-10 2 -4 3-5 2-4 co.5 z
liquid
region
MSRE fuel 11.1 6.1 11.1
(nominal)
Hastelloy N 69 16 I 5 -1
(nominal)
‘Semiquantitative spectrographic determination.

Table I 1.2. Gamma spectrographic (Ge-diode) analysis of deposits from mist shield in the MSRE pump bowl
Half-life 2.1 X 105 years 35 days
(after 9523)
Inventory, dis min-’ g-”
Sample activity per gram expressed
as fraction of MSRE inventory
activity/grams fuel saItb
Pump bowl bottom,
loose particles
ILatch stop
Top
Outside
Inside
Middle outside
Below liquid surface
Inside No. 1
Inside No. 2
Gage rod
Bottom
Outside
hide
99yc 9”Nb ‘03Ihl 106RU “‘Sb 127~1~~ ‘17cs 9529 144&
24 fig/g 8.3 ElQ
463
39.6 days
3.3 El0
367 days 2.7 years 105 days 30 years 65 days
3.3 E9 3.1 E8 2.0 E9 6.2 I3 9.9 El0
0.23 36 52 20 17 2.1 0.13 -t 0.03
265 364 1000 5.7 69 3.1 0
122 167 328 104 98 9.5 0
42 ?r 14 237 f 163 273 1000 69 4.3 0
271 531 646 563 54 8.Q 0
354 164 224 271 143 0.6 0
148 292 310 72 98 1.1 0
305 221 692 198 181 a.2 0
(0, 401 1210 167 232 3.2 Q
15 + 9 189 51 27 0.32 0
‘Background values (limit ofdetection) were as follows: 95Zr, 2-9 ElO; 14”Ce, l-2 ElO; *34Cs, 2-9 E$; 1 “Ag, l-3 E9; “‘Eu, 1 Eg-2 E9.
bLJncertainty stated (as f value) only when an appreciable fraction (>lO%) of observed.
284 days
5.9 El0
(3.10
0
0
0
0
0.13 + 0.02
0
0
0
0.11

. >:*
-Y
-"
materials concentrated in the same proportion should
have similar values.
We first note that the major part of these deposits
does not appear to be fuel salt, as evidenced by low
values of 95Zr and 44Ce. The values 0.13 and 0.1 1 for
Ce average 1276, and this is to be compared with the
combined 24% for LiF plus BeF, determined spectrographically,
as noted above. These would agree well
if fuel salt had been occluded steadily as 24% of a
growing deposit throughout the operating history.
For 7Cs we note that samples below liquid level
inside generally are below salt inventory and could be
occluded fuel salt, as considered above. For samples
above the liquid level inside, or any external sample,
values are two to nine times inventory for fuel salt.
Enrichment from the gas phase is indicated. Houtzeel'
has noted that off-gas appears to be returned to the
main loop during draining, as gas from the drain tanks is
displaced into a downstream region of the off-gas
system. However, our deposit must have originated
from something more than the gas residual in the pump
bowl or off-gas lines at shutdown. An estimate substantpating
this is as follows.
With full stripping, 3.3 X 18'
atoms of the mass 13'9
chain per minute enter the pump bowl gas. About half
actually go to off-gas, and most of the rest are
reabsorbed into salt. If we, however, assume a fraction f
is deposited evenly on the boundaries (gas boundary
area about 16,000 crn2), the deposition rate would be
about 2 x IO' 3f atoms of the 137 chain per square
centimeter per minute. Now if in our samples the
activity is I relative to inventory sait (I .4 x 10' ' atoms
of 'Cs per gram) and the density is about 2, then the
time t in minutes required to deposit a thickness of X
centimeters would be
In obtaining our samples, we generally scraped at least
0.1 g from perhaps 5 em" indicating a thickness of at
least about 0.81 cm, and 1 values were about 4, whence
tis about 6OO/f.
Thus, even if all
- 1) the 137 chain entering the
pump bowl entered our deposits, 600 min of flow
would be required to develop their '"Cs content -
too much for the 7-min holdup of the pump bowl or
even the rest of the off-gas system, excluding the
charcoal beds.
It appears more likely that 76s atoms, from 'Xe
atoms decaying in tke pump bowl, were steadily
incorporated to a slight extent in a slowly growing
deposit.
The noble-metal fission products, 35-day 'I Nb,
39.6-day ' O4 Ru. 367-day ' O6 RLI>
2.7-year 25 Sb, and
105-day 27mTe, were strongly present in essentially
all samples. In dl cases, 35-day '§i\jb was present in
quantities appreciably more than could have resulted
from decay of Zr in the sample.
Antimony-125 appears to be strongly deposited in all
regions, possibly more strongly in the upper (gas) region
deposits. Qearly ' "Sb must be considered a noble-
metal fission product. Tellurium-1 27rn was also found,
in strong concentration. frequently in similar propor-
tion to the "'Sb of the sample. 'Phe precursor of
127m
Te is 3.9-day ' 27Sb. It may be that earlier
observations about fission product tellurium are in fact
observations of precursor antimony isotope behavior,
with tellurium remaining relatively Fled.
The ruthenium isotopes were present in quantities
comparable with those of 95h%9 '"Sb, and 127naTe.
If the two ruthenium isotopes had been incorporated in
the deposit soon after formation in the sdt, then they
should be found in the same proportion to inventory.
But if a delay or holdup occurred, then the shorter-lived
Ru would be relatively richer in the holdup phase as
discussed later, the activity ratio ' ' Ku/l O6 Ru would
exceed the inventory ratio, and material deposited after
an appreciable holdup would have an activity ratio
' Ru/' Ru which would be less than the inventory
vdue. Examination of Table 11.2 shows that in all
samples, relatively less ' O RU was present, which
indicates that the deposits were accumulated after a
holdup period. This appears to be equally true for
regions above and below the liquid surface. Thus we
ond dude that the deposits do not anywhere represent
residues of the material held up at the time of
shutdown but rather were deposited over an extended
period on the various surfaces from a common holdup
source. Specifically this appears true for the lumpy
deposits on the mist shield interior and cage rods below
the liquid surface.
Data for 2.1 X 105-year 99Tc are available for one
sample taken from the inner mist sl&ld surface below
the liquid level. n e vaiue, 1.11 x lo4 pg/g, vs
inventory 24 yg/g, shows an enhanced concentration
ratio sindar to our other noble-metal isotopes md
dearly substantiates the view that this element is to be
regarded as a noble-metal fission product. The consistency
of the ratios to inventory suggests that the
noble metals represent about 5% of the deposits.
The quantity of noble-metal fission products held up
in this pump bowl fdm may not be negligible. If we
take a median value sf about 300 times inventory per
gram for the deposited material, take pump bowl area
i~ the gas region as 1 O,OOO cm2 (minimum), and assume

deposits 0.1 mm thick (about 0.02 g/cmz ; higher values
were noted), the deposit thus would have the equivalent
of the content of more than 60 kg of inventory salt.
There was about 4300 kg of fuel salt; so on this basis,
deposits containing about 1.4% or more of the noble
metals were in the gas space. At least a sirnilar amount
is estimated to be on walls, etc., below liquid level; and
no account was taken for internal structure surfaces
(shed roof, deflector plates, etc., or overflow pipe and
tank).
Since pump bowl surfaces appear to have more (about
10 times) noble-metd fission products deposited on
them per unit area than the surfaces of the heat
exchanger, graphite, piping, surveillance specimens, etc.,
we believe that some peculiarities of the pump bowl
environment must have led to the enkanced deposition
there.
We first note that the pump bowl was the site of
leakage and cracking of a few grams of Bubricating oil
each day. Purge gas flow also entered here, and
hydrodynamic conditions were different from the main
loop.
The pump bowl had a relatively hgh gas-liquid
surface with higher agitation relative to such surface
than was the case fOr gas retained as bubbles in the
main loop. The liquid shear against walls was rather less,
and deposition appeared thickest where the system was
quietest (cage rods). The same material appears to have
deposited in both gas and liquid regions, suggesting a
common source. Such a source would appear to be the
gas-liquid interfaces: bubbles in the liquid phase and
droplets in the gas phase. It is krmown that surface-
seeking species tend to be concentrated on droplet
surfaces.
The fact that gas and liquid samples obtained in
capsules during operation had ' RU/' ' RU activity
ratios higher than inventory and deposits discussed here
had ' RuI1 ' Ru activity ratios below inventory
suggests that the activity in the capsule samples was
from 3 held-up phase that in time was deposited on the
surfaces which we examined here.
The tendency to aggIomerate and deposit in the less
agitated regions suggests that the overflow tank may
have been a site of heavier deposition. The pump bowl
liquid which entered the overflow pipe doubtless was
associated with 3 high proportion of surface, due to
rising bubbles; this would serve to enhance transport to
the overflow tank.
The binder material for the deposits has not been
established. Possibilities include tar material and per-
haps structural- or noble-metal colloids. Unlikely,
though not entirely excludable, contributors are oxides
120
formed by moisture or oxygen introduced with purge
gases or in maintenance operations. The fact that the
mist shield and cage were wetted by salt suggests such a
possMity.
11.2 Exmhation of Moderator Graphite
from MSRE
A complete stringer of graphite (located in an add
position between the surveillance specimen assembly
and the control rod thimble) was removed intact from
the MSRE. TMs 64.5in.4ong specimen was transferred
to the hot cells for examination, segmenting, and
sampling.
1 B 2.1 Results of visual examination. Careful exami-
nation of dl surfaces of the stringer with a Koumorgen
periscope showed the graphite to be generally in very
good condition, as were the many surveillance speci-
inens previously examined. The corners were clean and
sharp, the faint circles left upon milling the fuel channel
surfaces were visible and appeared unchanged, and the
surfaces, with minor exceptions described below, were
dean. "he stringer bottom, with identifying letters and
numbers scratched on it, appeared identical to the
photograph taken before its installation in MSm.
Examination revealed a few solidified droplets of
flush salt adhering to the paphite, and one small
@.5-cm2) area where a grayish-white material appeared
to have wetted the smooth graphite surface. One small
crack was observed near the middle of a fuel channel.
At the top surface of the stringer a heavy dark deposit
was visible. This deposit seemed to consist in part of
salt and in part of a carbonaceous material; it was
sampled for both chemical and radiochemical analysis.
Near the metal knob at the top of the stringer a crack in
the graphite had permitted a chip (about 1 mrn thick
and 2 cm2 in area) to be cleaved from the flat top
surface. Ths crack may have resulted from mechanical
stresses due to threading the metal knob inlo the
stringer (or from thermal stresses in this metal-graphite
system during operation) rather than from radiation or
chemical effects.
1 B -2.2 Segmenting of gaphite stringer. Upon corn-
pletion of the viswal observation and photography and
after removal of small samples from several locations on
the surface, the stringer was sectioned with a thin
Carborundum cutting wheel to provide five sections of
4-in. length, three thin (10- to 28-mil) sections for x
radiography, and three thicker (38- to 6O-mil) sections
for pinhole scanning with the gamma spectrometer.
Each set of samples contained specimens from near the
top, middle, and bottom of the stringer. The large
k. .&

121
specimens, from which surface samples were subse- graphite stringer using a rotating milling cutter 0.619 in.
quently milled, included (in addition) two samples from in diameter. The specimen was clamped flat on the bed
intermediate positions.
of the machine, md cuts were made from the flat
11.2.3 Examination of surface samples by x-ray fuel-channel surface and from one of the narrower flat
diffraction. Previous attempts to determine the chermi- edge surfaces on both sides of the fuel channel. The
cal form of fission products deposited on graphite latter surfaces contacted the similar surfaces of an
surveillance specimens by x-ray reflection from flat adjacent stringer in the MSWE core. After the sample
surfaces failed to detect any element except graphitic was clamped in position the milBing machine was
carbon. A sampling method urhicb concentrated surface operated remotely to take successive cuts 1, 2, 3, 10,
impurities was tried at the suggestion of Harris Dunn of 20, 30, 245, 245, and 245 mils deep to the center of
the Analytical Chemistry Division. This method in- the graphite stringer. The powdered graphite was
volved lightly brushing the surFace of the graphite collected in a tared falter bottle attached to a vacuum
strbger with a fie Swiss pattern fide which had a cleaner hose during and after each cut. The filter bottle
curved surface. The grooves in the file picked up a small was a plastic cylindrical bottle with a circular falter
amount of surface material, which was transferred into paper covering a number of drilled holes in the bottom.
a glass bottle by tapping the fde on the lip of the bottle. This technique collected most of the thinner samples
In this wayy, samples were taken at the top, middle, and but only about half of the larger 245-mil samples. After
bottom of the graphite stringer from the fucl-channel each sampling, the uncollected powder was carefully
surface, from the surface in contact with graphite, and removed with the empty vacuum cleaner hose.
from the curved surface adjacent to the control rod Before samples were cut from the narrow flats the
thimble.
corners of the stringer bars were milled off to a width
Three capillaries were packed and mounted in holders and depth of 66 mils to avoid contamination from the
which fitted into the x-ray camera.
adjacent stringer surfaces. Then successive cuts 1, 2, 3.
Samples from the fuel-channel surfaces yielded very 10, 20, and 30 mils deep were taken. Finally, a single
dark fiims, which were difficult to read. Many weak cut 62 mils deep was taken on the opposite Eat
lines were obsemed in the x-ray patterns. Since other fuel-channel surface. Only the latter cut was taken on
analyses had shown Mo, Te, Ru, Tc, Ni, Fe, and Cr to the two stringer samples from positions halfway bebe
present in significant concentrations on the graphite tween the bottom and middle and halfway between the
surface, these elements and their carbides and tellurides middle and top of the stringer.
were searched for by careful comparison with the 11.2.5 Ra$iocheanical and chemical analyses of
observed patterns.
In three of the graphite surface samples andyzed,
MSRE graphite. The milled graphite samples were
dissolved in a boiling mixture of concentrated H2So4
most of the lines for MonC and ruthenium metal were and HN03 with provision for trapping any volatilized
certainly present. For one sample, most of the lines for activities. The following analyses were run on the
Cr,C3 were seen. (The expected chromium carbide in dissolved samples:
equilibrium with excess graphite is Cr3 Cn , but nearly
half the diffraction lines for this compound were
missing, including the two strongest lines.) Five of the
six strongest lines for NiT'ee, were observed. Molybdenum
metal, telluriuni metal, technetium metal, chromium
metal, CrTe, and MoTe2 were excluded by
I. Gamma spectrometer scans for lQ6~u, 125~b,
134Cs,'37Cs,1 "Ag, 144Ce,g55r,""b,and 6oCo.
2. Separate radiochemical separations and analyses for
"Sr, "Sr, lz7Te, and 'H. A few samples were
andyzed for E.
comparison of their known pattern with the observed
spectrum. These observations (except for that of
Cr&) are in accord with expected chemical behavior
and are significant in that they represent the first
experimental identification of the chemical. form of
fission products known to be deposited on the graphite
3. Uranium analyses by both the fluorometric and the
delayed neutron counting methods.
4. Spectrographic analyses for E, Be, Zr, Fe, Ni, Mo,
and Cr. (Highyield fission products were also looked
for but not found.)
surface.
Uranium and spectrographic mdyses. Table 11.3
11.2.4 Milling ~f surface graphite samples. Surface gives the uranium andyses (cdculated as 33~) by both
samples for chemical md radiochemical analyses were the fluorometric method mmd the delayed neutron
nfled from the five 4-h-long segments from the topt counting method. The type of surfaces sampled, the
middle, bottom, and two intermediate locations on the number of the cut, and the depth of the cut for each

Smpk
1
2
3
4
5
7
11
12
13
14
14
18
19
20
21
23
26
27
28
29
31
32
33
34
35
36
39
43
44
45
48
49
50
Cut and
typeb
Depth,
mils
Total U, ppm,
fluorometric
a33U, PPm
delayed
neutron
Li, wm Be, P P ~ Zr, ppm
Metal,C
PPm

... . .
v;.;@
values indicate that uranium penetration into modera-
tor graphite should not be a serious problem in
large-scale molten-salt reactors.
The fact that fluorometric values for total uranium
and the delayed neutron counting values for 2 4 3 ~
agreed (with the 233U vdue usually larger than the
total uranium value) indicates that little uranium
remained in the graphite from the operation of the
MSWE with U fuel. Apparently, the * U and U
previously in the graphite underwent rather complete
isotopic exchange with 233U after the fuel was
changed. The finding of 150 to 2905 ppm nickel in a
few of the surface samples is probably red, The main
conclusions from the spectrographic analyses are that
adherent or permeated fuel salt accounted for the
uranium, lithium, and beryllium in half the sarnples and
that a thin layer of nickel was probably deposited on
some of the graphite surface.
Radiochemicd analyses. Since the graphite stringer
samples were taken more than a year after reactor
shutdown, it was possible to analyze only for the
reiatively long-lived fission products. However, the
absence of interfering short-lived activities made the
analyses fur long-lived nuclides more sensitive and
precise. The radiochemkal analyses are gven in Table
11.4, together with the type and location of surface
sampled, the number of the milling cut, and the depth
of the cut foe each sample.
The species IZ5Sb, '"Ru, 1'oAg,"Nb,and"271e
showed concentration profiles similar to those observed
for noble-metal fission products (9 Mo, Te, ' Te,
' Ru. ' Ru, and ' Nb) in previous graphite sui~ed-
lance speciniens, that is, high surface concentrations
fang rapidly several orders of magnitude to low
interior concentrations. he profiles for ~r were of
similar shape, but the interior concentrations were two
or three orders of magnitude smaller than for its
daughter ' ~ b IB . is thought that Zr is either injected
into the graphite by a fission recoil rnech'anisaism or is
carried with fuel salt into cracks in the graphite; the
much larger concentrations of "h% (md the other
noble metals) are thought to result from the deposition
or plating of solid metallic or carbide particles on the
graphite surface. The "Sr (33-sec "Kr precursor)
profiles were much steeper than those previously
observed in surveillance specimens for "~r (3.2-min
88fi precursor), as expected. An attempt to analyze
the stringer samples for Sr also was unsuccessful. It is
difficult to malyze for one of these pure beta emitters
in the presence of Barge activities of the other.
Surprisingly high concentrations of tritium were
found in the moderator gaphite samples (Table 11 -4)-
123
Re tritium concentration decreased rapidly from about
10" dis niiii-' g-' at the surface to about IO9 dis
miK' g-' at a depth ui 'jI6 in. and then decreased
slowly to about half this value at the center of the
stringer. If all the graphite in the MSRE contained this
much tritium, then about 15% of the tritium produced
during the entire power operation had been trapped in
the graphite. About half the total trapped tritium was
in the outer '/' 6-in. layer.
Similarly high concentrations of tritium were found
in specimens of Poco graphite (a graphte characterked
by large uniform pores) exposed to fissioning salt in Bhe
core during the find 1786 hr of operation. Surface
concentrations as Iii& as 4.5 x IO" dis rnin-' g-'
were found, but interior concentrations were below I O8
clis min-' g-', much lower than for the moderator
graphite. This suggests that the graphite surface is
saturated relatively quickly but that diffusion to the
interior is slow.
If it is assumed that the surface area of the graphite
(about 0.5 m2/g) is not changed by irradiation (there
was no dependence of tritium sorption on flux), there
was 1 tritium per 100 surface carbon atoms. Since the
MSRE cover gas probably contained about 100 times as
much hydrogen (from pump oil decomposition) as
tritium, a remarkably complete coverage by chenaisorbed
hydrogen is indicated.
An overall assessment of tritium behavior in the
MSWE3 and proposed MS14R's4 is presented elsewhere.
The data on fission product deposition on and in the
gtaphite, based on Table 1 E A, hdve been calculated as
(observed activity per square centimeter) divided by
(inventory activity/total area), as was done for surveillance
specimens earlier. The resulting relative deposit
intensities, shown in Table 1 B 5, can, of co~erse, then be
compared with values reported fur other nuclides, other
specimens, or other times. If dl graphite suxfaces were
evenly covered at the indicated intensity, the fraction
of total inventory in such deposits would be 74% of the
relative deposit intensity shown. For top, middle, and
bottom regions and for channel and edge (graphite-tographite)
surfaces, values are ~hwn for surface and
overall deep cuts.
As in the case of surveillance specimens, intensities
for salt-seeking nuclides are at levels appropriate for
fission recoil (about 0.0511, the noble-gas daughters
(I 7Cs and "Sr) are 18 to 20 times as high, and the
noble metals notably higher, though, as we shall see,
not as high on balance as on metals. Mere comparisons
can be made, most of the deposit was indicated to be at
the surface. Values for 9 5 are ~ far above 95~r,
indicating that N~I was indeed deposited; the graphite

.._...-.-.
Table B 1.4. Radiochemical analyses of graphite stringer samples
Sample cut and Depth, - -_ Disintegrations per minu$ per gram of graphite on 12-12-69
No. type’” mils ‘2SSb ro6RU “27Te 95iVb 1 lOAg 99gc 134cs 137cs 9% lf14Ce 95Zr IS2EU l S4EU 6OCO 3H
1
2
3
4
5
6
7
8
9
10
II
12
18
19
20
22
23
24
26
27
28
36
31
32
33
34
35
38
40
42
43
44
49
50
51
MD
Thin blank 0 6

. . .,+, . . :. ..X.b . . . . .-..;
kc.&/
125
Table 11.5. Fission pKdUCtS in MSRE graphite COR bar after removal in cumulative values of ratio to inventory
--
I____-_ ~
Location Type Depth
(mils)
1 3 7 ~ ~ gos, 1 4 4 ~ ~ g5zZr 9sm I O 1~ 2 7 ~ ~ ~ ~
TOP Channel 0-2 0.0008 0.007 0.0007 0.24a
0-800 0.087 0.072 0.0032 0.0024 0.69
Edge 0-2 0.0008 0.006 0.0005 0.0007 0.12
Edge 0 62 0.0007 0.038 0.30
Middle Channel 0-3 0.0029 0.030 0.002 0.31
0-550 0.071 0.003
Edge 0 3 0.013 0.014 0.0008 0.0008 0.15
0-36 0.020 0.029 0.0011 0.0009 0.26
Edge 0 62 0.030 0.075 0.0004 1 .oo
Fdge 0-62 0.037 0.081 0.0021 0.0023 0.53
Bottom Channel 0-3 O.OQ15 0.014 0.0012 0.0011 0.43
0 800 0.069 0.017 0.0024 0.0011 0.58
Edge 0-3 0.0017 0.006 0.0015 0.0015 0.45
‘Includes signlficant Q values.
values appear to be higher than those reported in the
next section for metal surfaces, but the passage of a
dozen half-lives between shutdown and determination
may have reduced the precision of the data.
11 .S Examination of Heat Exchangers and Control
Rod mmMe surfaces
Among the specimens of component surfaces excised
from the MSRE were two segments of control rod
thimble and one each of heat exchanger shell and
tubing. In addition to the fission product data we
report here. these specimens were of major interest
because of grain-boundary cracking of surfaces contacting
molten salt fuel, as discussed elsewhere.5
Successive layers were removed electrochemically
from the samples using methanol--l0% HCl as electrolyte.
These solutions were examined both spectrographically
and radiochemically. The sample depth was
calculated from the amount of nickel and the specimen
geometry, etc. Concentration profiles (relative to
nickel) for the fuel side of the specimen of heat
exchanger tube are shown in Fig. 11 -5 €or constituent
elements. fission product tellurium, and various fission
product nuclides. It is of interest to note that concentrations
3 mils below the surface were 1 to 10% of
those observed near the surface. Values this high could
Inventory (di%/inm per gram of salt)
6.2E9 6.169 5.9EIO 9.9E10 8.3E10
0.11 0.024
0.14 0.090
0.037 0.036
0.38 0.68
0.05 0.10
0.12
0.021 0.015
0.024 0.018
1.16 0.94
0.29 0.61)
0.09 0.36
0.14 0.43
0.07 0.14
3.3E9 2.OE9
iasSb
0.41
0.50
0.054
0.79
0.1 1
0.24
0.25
2.14
0.79
0.68
0.84
0.27
3.7E8
be the result of a grain-boundary transport of the
nuclides.
For the various samples, deposit concentrations were
obtained by summing the amounts determined in
successive layers. These are shown, expressed as ratios
to inventory concentration divided by total MSRE area
(3.0 X IO6 cm2)>, in Table 11.6.
As discussed for surveillance specimen and other
deposit data, this ratio, the relative deposit intensity,
permits direct comparison with deposits on other
surfaces. for example, graphite. To calculate the fraction
of inventory that deposits on a particular kind of
surface, the relative deposit intensity should be multiplied
by the fraction of MSRE surface represented by
the deposit. The metal surface was 265% of the total
MSRE surface.
The most strongly deposited nuclides were Sb and
27Te, with most values above I, indicative of a
selective strong deposition. The variation in values from
surface to surface suggests that generalizing any value to
represent all metal surfaces will probably not be very
accurate. Also. no values are available for certain large
areas -- in particular, the reactor vessel surfaces. In spite
of all these caveats, the data here for tellurium and
antimony are consistent with similar data from the
various surveillance specimens in indicating that these

elements are very strongly deposited on metal surfaces,
as well as somewhat less strongly on graphite.
It is sufficient here to note that less strong but
appreciable deposition of ruthenium? technetium, and
niobium was found.
Fig. 11.5. Concentration profiles faom the fuel side of an
MSRE heat exchanger tube measured about 1.5 years after
reactor S~U~~QWII. (Arrows indicate Ievel was less than or equal
to that given.)
11.4 Metal Transfer in MSWE Salt Circuits
Cobalt-60. is formed in Hastelloy N by neutron
activation of ihe minor amount of Co (O.SS%> put in
the alloy with nickel; the detection of 6oCo activity in
bulk mepal serves as a measure of its irradiation history,
and the detection of "Co activity on surfaces should
serve as a measure of metal transport from irradiated
regions. Cobalt-BO deposits were found on segments of
coolant system radiator tube, on heat exchanger tubing,
and on core graphite removed from the MSRE in
January I97 1.
The activity found on the radiator tubing (which
received a completely negligible neutron dosage) was
about 160 dis inin-' cm-'. This must have been
transported by coolant salt flowing through heat
exchanger tubing activated by delayed neutrons in the
fuel salt. The heat exchanger tubing exhibited subsurface
activity of about 3.7 X IO3 dishin per cubic
centimeter of metal, corresponding to a delayed neutron
flux in the heat exchanger of about I x 10' '. ~f
metal were evenly removed from the heat exchanger
and evenly deposited on the radiator tubing throughout
the history of the MSRE, a metal transfer rate at fun
puwei of about 0.0005 millyear is indicated.
Cubalt-60 activity in excess of that induced in the
heat exchanger tubing was found on the fuel side of the
tubing (3.1 X 106 dis niiri-' cm-?-) and on the samples
of core graphite taken from a fuel channel surface (5 X
to 3.5 X
dis min-' cmC). he higher values
on the core graphite and their consistency wit11 iluence
imply that additional activity was induced by core
neutrons acting on "Co after deposition on the
graphite.
The reactor vessd (and annulus) walls are the major
metal regions subject to substantial neutron flux. If
these served as the major source of transported metal
and if this metal deposited evenly on all surfaces, a
metal loss rate at full power of about (4.3 xnilj'year is
indicated. Because deposition occurred on both the
hotter graphite and cooler heat exchanger surfaces.
simple thermai transport is not indicated. Thermodynanzic
arguments preclude oxidation by fuel.
One mechanism for the indicated metal transport
might have 10% of the 1.5 W/cm3 fission fragment
en erg^ in the annular fuel within a 30-p range deposited
in the metal and a small fraction of the metal sputtered
into the fuel. About 0.4% of the fission fragment
energy entering the metal resulting in such transfeet
would correspond to the indicated reactor vessel loss
rate of 0.3 miliyear. If this is the correct mechanism,
reactors operating with higher fuel power densities

.=y
adjacent to metal should exhibit proportionately higher
loss rates.
1 1.5 Cesin~n Isotope Migation in MSE Graphite
Fission product concentration prof& were obtained
on the graphite bar from the center of the MSRE core
which was removed early in 1971. The bar had been in
the core since the beginning of operation; it thus was
possible to obtain profiles for 2.1-year ' 34Cs (a
neutron capture product of the stable Cs daughter
of 5.27-day 33~e) as well as the 3O-year ' 3 7 ~ s
daughter of a.9-min 7Xe. These profies, extending
to the center of the bar, are shown in Fig. 11.6.
Graphite surveillance specimens exposed for shorter
periods, and of thinner dimensions, have revealed
similar profiles for CS.~ Some of these, along with
profiles of other rare-gas daughters, were used in an
analysis by Ked7 of the behavior of short-lived noble
gases in graphite. Xenon diffusion and the possible
formation of cesium carbide in molten-salt reactors
have been considered by Baes and Evans.'
An appreciable literature on the behavior of fission
product cesium in nuclear graphite has been developed
in studies for gas-cooled reactors by British kvestigators,
the Dragon Project, Gulf General Atomic
workers, and workers at ORNL.
The profiles shown in Fig. 11.6 indicate significant
diffusion of cesium atoms in the graphite after their
formation. The 5.24-day half-life of 33Xe must have
resulted in a fairly even concentration of this isotope
125
shows that this occurred. The ' Cs concentration that
would accumulate in graphite if no diffusion occurred
has been estimated from the power history of the
MSRb to be about 2 X atoms per gram of
graphite (higher if pump bowl xenon stripping is ._ -
inefficient), assuming the local neutron flux was a
minimum of four times the average core flux. The
observed ' 34Gs concentration in the bar center, where
diffusion effects would be least, was about 5 X 10l4
atom of ' 3 4~s
per gram of graphite. The agreement is
not unreasonable.
Data for N-year ' 7Cs are shown in Fig. I 1.6. c or
comparison, the accumulated decay profile of the
parent 3.9-miti 7Xe is shown as a dotted line in the
(u to,7
throughout the graphite and must have produced a flat 0 100 200 300 400 500 660 706 800
deposition profile for 'Cs. This isotope and its DEPTH (mils)
neutron product 34Cs could diffuse to the bar surface
and could be traken up by the sdt. The CS profile
to15
11.6. concentPation of cesium isotopes in MSRE ore
graphite at given distances from fuel c~mnei
surface.
Surface inventory !35m 99Tc 1WRU 1OSRu 127Te 12ssb
- ~~ ~~ ~ ~ ~
Control rod thimble, bottom 0.14 1.2 1.5 0.5Q 1 .6§ 3.3
Control rod thimble, middle 1.0 0.73 0.58 0.42 0.51 1.4
Kist shield outside, liquid 0.26 0.73 0.27 0.38 0.89 2.8
Heeat exchanger, shell 8.33 1 .0 0.10 0.1 9 1.4 2.6
Heat exchanger. tube 0.27 1.2 0.1 I 0.54 2.6 4.3
MSRE inventmy divk-kd by total MSRE surface area (dis min-' cm -')
8.3E10 9E5 3.3EIO 3.3E9 2.OE9 3.7B8
-

upper left of the figure. TBtis was estimated assuming
that perfect stripping occurred in the pump bowl with a
mass transfer coefficient from salt to central core
graphite of 0.3 ftjhr' and a diffusion coefficient of
xenon in graphite (10% porosity) of 'I X
cm21sec.' '
Near the surface the observed 13'Cs profile is lower
than the estimated deposition profile ; toward the center
the observed profile tapers downward but is about the
es imated deposition profile. This pattern should develop
if diffusion of cesium occurred. The central
concentration is about one-third of that near the
surface. Steady diffusion into a cylinder' from a
constant surface source to yield a similar ratio requires
fiat Dt/r2 be about 0.14. For a cylinder of 2 crn radius
and a salt circulation time of 21,788 hr, a cesium
diffusion coefficient of about 7 X IO-' cm2/sec is
indicated.
Data developed for cesium-in-graphite relationships in
gas-cooled reactor systems' at temperatures of $00 to
t100"C may be extrapolated to 650°C for comparison.
The diffusion coefficient thereby obtained is slightly
below IO-' '; the diffusion coefficient for a gas (xenon)
is about 1 x cm2/sec. Some form of surface
diffusion of cesium seems indicated. This is further
substantiated by the sorption behavior reported by
Milstead' for the cesium-nuclear-graphite system.
In particular, Milstead has shown h t at temperatures
of 880 to 1 100°C and concentrations of 0.04 to 1.6 mg
of cesium per gram of graphite, cesium sorption on
graphite follows a Freundlich isotherm (1.6 mg of
cesiuni on 1 m2 of graphire surface corresponds to the
saturation surface compound CsC8). Below this, a
hngmuir isotherm is indicated. ]in the MSRE graphite
under consideration the "Cs content was about 10'
atoms per gram of graphite, and the 133 and 135
chains would provide similar amounts, equivalent to a
total content of 0.007 rng of cesium per gram of
graphite. At 65O"C, in the absence of interference from
other adsorbed species, Eangmuir adsorption to this
concentration should occur at a cesium partial pressure
of about 2 x IO-'* atm. At this pressure, cesium
transport via the gas phase should be negfigible, and
surface phenomena should control.
To some extent, rubidium, strontium, and barium
atoms also are indicated to be similarly adsorbed and
likely to diffuse in graphite.
It thus appears that for time periods of the order of a
year or more the alkali and alkaline earth daughters of
noble gases which get into the graphite can be expected
to exhibit appreciable migration in the moderator
graphite of molten-salt reactors.
B 28
11 -6 Noble-Metal Fission Transport Model
It was noted elsewhere' that noble metals in MSIPE
salt samples acted as if they were particulate con-
stituents of a mobile "pool" of such substances held up
in the system for a substmtial period and that evidence
regarding this might be obtained from the activity ratio
of pairs of isotopes.
Pairs of the same element, thereby having the same
chemical behavior (e.g., Ru and lo Ru), should be
particularly effective. As produced, the activity ratio of
such a pair is proportional to ratios of fission yields and
decay constants. Accuinulatisn over the operating
history yields the inventory ratio, ultimately propor-
tional (at constant fission rate) only to the ratio of
fission yields. If, however, there is an intermediate
holdup and release before final deposition, the activity
ratio of the retained material will depend on holdup
time and will fall between production and inventory
values. Furthermore, the material deposited after such a
huldup will, as a result, have ratio values lower than
inventory. (Values for the isotope of shorter half-life,
here 03Ru, will be used in the numerator of the ratio
throughout our discussion.) Consequently, comparison
of an observed ratio of activities (in the same sampk)
with associated production arid inventory ratios should
provide an indication of the "accumulation history" of
the region represented by the sample. Since both
determinations are for isotopes of the same element in
the same sample (consequently subjected to identical
treatment), many sampling and handling errors cancel
and do nut affect the ratio. Ratio values are thereby
subject to less variation.
We have used the ' ' Wu/' ' Ru activity ratio, among
others, to examine samples of various kinds taken at
various times in the MSRE operation. These include salt
and gas samples from the pump bowl and other
materials briefly exposed there at various times.
Data are also available from the sets of surveillance
specimens removed from runs 11, 14, 18, and 20.
Materials removed from the off-gas line after runs 14
and I8 offer useful data. Some information is avail-
able' from the on-site gamma spectrometer surveys of
the MSWE following runs 18 and 19, particularly with
regard to the heat exchanger and off-gas line.
11.6.1 Inventory and model. The data will be dis-
cussed in terms of a "compartment" model, which will
assign first-order transfer rates common for both
isotopes between given regions and will assume thar this
behavior was consistent throughout MSKE history.
Because the half-lives of ' 03Ru and 06Ru are quite
different, 39.6 and 367 days, respectively, appreciably
~ . ~ .
., . . . . . .
. .. .. .

different isotope activity ratios are indicated for differ-
ent conipartmcnts and tinies as simulated operation
proceeds. A sketch of a useful scheme of compartments
is shown in Fig. 1 1.9.
We assume direct production of ’ ’ Ru and ’‘ Ru in
tlie fuel salt in proportion to fission rate and fuel
composition as determined by MSRE history. The
material is fairly rapidly lost from salt either to
“surfaces” or to a mobile “particulate pool” of agglom-
erated material. The pool loses material to one or more
final repositories, nominally “off-gas,” and also may
deposit materid on the “surfaces.” Rates are such as to
result in an appreciable holdup period of the order of
50 to 100 days in the “particulate pool.” Decay, of
course, occurs in all compartments.
Material is also transferred to the “drain tank” as
required by the history, and transport between com-
partments ceases in the interval.
From the atoms of each type at a given time in a
given compartment, the activity ratio can be calculated,
as well as an overall inventory ratio.
We shall identify sarnples taken from different regions
of the MSRE with the various compartments and thus
obtain insight into the transport paths and lags leading
to the sampled region. It should be noted that a
compartment can involve more than one region or kind
of sample. The additional information required to
establish the amounts of material to be assigned to a
given region, and thereby to produce a material balance,
is not available.
(AND HEEL)
In comparison with the overall inventory value of
’ Ru,” ’ Ru, we should expect “surface” values to
equal it if the deposited rnaterial comes rapidly and
only from “salt” and to be somewhat below it if, in
addition. “‘particulate” is deposited. If there is no direct
deposition from “salt” to “surface,” but only “particu-
late,” then deposited material should approach “off-
gas” compartment ratio.
The “off-gas” compartment ratio should be below
inventory. since it is assumed to be steadily deposited
from the “particulate pool,” which is richer in the
long-lived O6 Ru component than production, and
inventory is the accumulation of production minus
decay.
The particulate pool will be above inventory if
material is txansferred to it rapidly and lost from it at a
significant rate. Slow loss rates correspond to long
holdup periods, and ratio values tend toward inventory.
Differential equations involving proposed transport,
accumulation, and decay of Io3Ru and ‘06Wu atoms
with respect to these compartments were incorporated
into a fourth-order Runge-Kutta numerical integration
scheme which was operated over the full M§RE power
history.
The rates used in one calculation referred to in the
discussion below show rapid loss (less than one day)
from salt to particulates and surface, with about 4%
going directly to surface. Holdup in the particulate pool
results in a daily transport of about 2% per day of the
pool to “off-gas,” for an effective average holdup
(ALL TRANSPORT CEASES DURING
PERIOR THAT SALT IS DRAINED
FROM SYSTEM)
Fig. 11.9. Compartment model for noblemetal fission transport in MSRE.

period of about 45 days. All transport proccsses are
assumed irreversible in this scheme.
11.6.2 Off-gas line deposits. Data were reported in
Sect. 10 on tlie examination dfter run 14 (March 1968)
of the jumper line installed after run 9 (December
19661, on the examination after run 18 (June 1969) of
parts of a specimen holdei assembly from the main
off-gas line installed after run 14, and on the exantination
of parts of line 523, the fuel pump overflow
tank purge gas outIet to the main off-gas line, which
was installed during original fabrication of MSRL
These data are shown in Table I H .7.
For the jumper line removed after run 14, observed
ratios range from 2.4 to 7.3. By comparison the
inventory ratio for the net exposure interval was 12.2.
If a holdup period of about 45 days prior to deposition
in the off-gas line is assumed, we calculate a lower ratio
for the compartment of 7.0. It scems indicated that a
holdup of ruthenium of 45 days or more is required.
Ratio values from the specimen holder removed from
the 52% line after run 18 ranged between 9.7 and 5.0.
Net inventory iatio for the period was 13.7. and for
material deposited after a 45-day holdup, we estimated
a ratio of 12.3. A longer holdup would reduce this
estimate. However, we recall that gas flow th~oqgh this
line was appreciably diminished during the final month
of run 18. This would cause the observed ratio values to
be lower by an appreciable factor than would ensue
from steady gas flow all the time. A holdup period of
something over 45 days still appears indicated.
Flow of off-gas through line 523 was less well known.
In addition to bubbler pas to measure salt depth in the
overflow tank, part of the main off-gas flow from the
pump bowl went through the overff ow tank when flow
through line 522 was hindered by deposits. The
observed ratio after run 18 was 8 to 13.7, inventory was
9.6, and fur material deposited after 45 days holdup the
ratio is calculated to be 5.8. However. the unusually
great flow during the final month of run 18 (until
blockage of line 523 on May 25) would increase the
observed ratio considerably. This response is consistent
with thc low value for material from line 522 cited
above. So we believe the assumption of an appreciable
holdup period prior to deposition in off-gas regions
remains valid.
H 1 A.3 Surveillance specimens. Surveillance specimens
of graphite and also selected segments of metal
were removed from the core surveillance assembly after
exposure throughout several runs. Table 11.8 shows
values of the activity ratios for 103Ru/'06Ru for a
number of graphite and metal specimens removed on
different occasions in 1967, 1968, and 1969. Insofar as
deposition of these isotopes occurred irreversibly md
with reasonable directness soon after fission, the ratio
values should agree with the net inventory for the
period of exposure, and the samples that had been
exposed longest at a given removal time should have
appropriately lower values for the ' RU/' au activ-
ity ratio.
Examination of Table 1 I .8 shows that this latter view
is confirmed - the older samples do have values that are
lower, to about the right extent. However, we also note
that most observed ratio values fall somewhat below the
net inventory values. This could come about if, in
addition to direct deposition from salt onto surfaces,
deposition also occurred from the Iioldup "pool,"
presumed colloidal or particulate, which was mentioned
in the discussion of the off-gas deposits. Few of the
observed values fall below the parenthesized off-gas
values. This value was calculated to result if all the
deposited material had come from the holdup pool.
Table 18.4. Ruthenium isotope activity ratios
of off-gas time deposits
Observed vs calculated
1 disimin Io3Ru
-. , ,
dis:min lob~u
Sample Observed Calculated
1. Jumper Sestion of Line 522, Exposed December 1966
to
7 r
March 25, 1968
Flextool
Productionb 58
Upstream hose
Net inventory 12.1
Downstream hose 4.1
Inlet dust i .3
Net deposi t if:
Ourlet dust
4.4 45-day holdup 7.0
90-day holdup 5.5
88. Specimen Bolder, Line 522, Exposed August 8968
I
eo June 1969
Rail end
Production' 43
Flake
Net inventory 19.7
Tube sections 5.4,s.g
Recount 7/70, corr 6,
z
2 Net deposit if 45-day holdup 12.3
90-day holdup 9.3
111. Overflow Tank Off-Gas Line (529, Exposed 1965
to June 1969
Valve V513
Line 523
Valve HCV 523
9.6
45-day holdup 5.8
90-day holdup 4.3
aCorrected to time of shutdown.
b435LJ-2381. fuel, with inbred 239Pu.
c2 LJ fuel, including 2.1 '%; of fissions from contained ' hi
and 4.3% from 239Pu. LA
L.&4

(9'6) P'E I L'IZ

____
I
._ e-
4
\
i
k
1
I
---
e
~
I
i s
-i B
.
K..ir>:.-.-
...
.,.

. *.- . . . . . . . :,.
. w.- Y
regions resulting in a limited retention rather than the
unlimited retention implied by an inventory value.
Although meaningful differences doubtless exist between
different kiilds of samples taken from the pump
bowl, their similarity clearly indicates that all are taken
from the same mobile pool, which loses material? but
slowly enough to have an average retention period of
several months.
Discussion. The data presented above repre3ent
practically ai1 the ratio data available for hfSRE
samples. The data based on gamma spectrometer
surveys of V ~ ~ ~ Qreactor U S regions, particularly after
runs 18 and 19. have not been examined in detail bait in
cursory views appear not too inconsistent with values
given here.
It appears possible to summarize our findings about
fission product suthenium in this way:
The off-gas deposits appear to have resulted from the
fairly steady accuniuiation of material which had been
retained elsewhere for periods of the order of several
months prior to deposition.
The deposits on surfaces also appear to have con-
tained material retained elsewhere prior to deposition,
though not tu quite the same extent, so that an
appreciable part could have been deposited soon after
fission.
All materials taken from the pump howl contain
ruthenium sot opes with a coinrnon attribute: they are
representative of an accumulation of several months.
Thus ali samples from the pump bowl presumably get
their ruthenium from a common source.
Since it is reasonable to expect fission products to
enter salt first as ions or atoms, presumably these
rapidly deposit on surfaces or are agglomerated. The
agglomerated material is not dissolved in salt but is
fairly well dispersed and may deposit on surfaces to
some extent. It is believed that regions associated with
the pump bowl -- the liquid surface, including bubbles,
the shed roof, mist shield, and overflow tank - are
effective in accumulating this agglomerated material.
Regions with highest ratio of salt surface to salt mass
(gas samples containing mist and surfaces exposed to
the gas-liquid interface) have been found to have the
highest quantities of these isotopes relative to the
amount of salt in the sample. So the agglomerate seeks
the surfaces. Since the subsurfrrce salt samples, however,
never show amounts of ruthenium in excess of in-
ventory, it would appear that material entrained,
possibly with bubbles. is fairly well dispersed when in
salt.
LOSS of the agglomerated material to one or more
permanent sinks at a rate of 1 or 2% per day is
indicated. In addition to the off-gas system and to some
133
extent the reactor surfaces, these sinks could include
the overflow tank and various nooks, crannies, and
crevices if they provided for a reasonably steady
irreversible loss.
Without additionai information the ratio method
cannot indicate how much material follows a particular
path to a particular sink. but it does serve to indicate
the paths and the transport lags dong them for the
isotopes under consideration.
References
I. E. L. Compere and E. G. Bohlmann. “Examination
of Deposits from the Mist Shieid in the MSRE Fuel
Pump Bowl,” MSR Program Semiannu. Prog. Rep. Feb.
28, 1971, pp. 76- 85.
2. A. Houtzeel, private communication.
3. P. N. Haubenreich, A Review of Production arid
Observed Distribution of Trifium in the MSRE in the
Light ojRecenF Findings, 8WNL-GI;-71-8-34 (Aug. 23,
1971). (Internal document - No further dissemination
authorized.)
4. W. B. Briggs, “Tritium in Molten Salt Reactors.”
Reactor Technol. 14,335 42 (Winter 1971-1972).
5. H. E. McCoy and B. hlcSabb, Intergranular Cpac k-
ing of IiVOR-8 iti the MSRE, QRNL-4829 (November
1972).
6. S. S. Kirslis et al., “Concentration Profiies of
Fission Products in Graphite,” MSR Prcgram SPmiannu.
Prop. Rep. Aug. 31, 1970, ORNL-4622, pp. 69 40;
Feb. 28, 1970, ORNL-4548, pp. 104 5;Fe’t.b. 28, 1967,
ORNL-4119, pp. 125-28;A~g. 31, 1966, OWSL-4037,
pp. 180 84 D. R. Cuneo et al., ‘“Fission Product
Profiles in Three MSRE Graphite Surveillance Speci-
mens,” MSR Program Swiiaanrzu. Progr. Rep. Azg. 31,
1968, 0RNL-4344, pp. 142, 144-47; Feh. 29, 2968,
ORNL-4254. pp. 116, 118- 22.
4 R. J. Kedl, A Model for Computing the kfigraFion
of Very Short Lived Noble Gases into MSKE Graphite,
ORNL-TM- 1 81 0 (July 1967).
8. C. E. Baes, Jr., and W. B. Evans III.MSR Program
Semiannu. Prop. Rep.p. AUK. 31, 1966, ORNL-4037, pp.
158-65.
9. R. Bo Briggs, “Estimate of the Afterheat by Decay
of Noble Metals in MSBR and Comparison with Data
Cram the MSRE.” MSR-68-138 (Nov. 4, 1968). (Inter-
nal document - No further dissemination authorized.)
10. R. B. Evans III, J. L. Rutherford, and A. P.
Malinauskas, Gas Transport in MSRE Moderator Graph-
ite> 11. Effects of Inipregiation, UI. Variation oj Fiow
Properties, ORSL-4389 (May 1969).
11. 3 Crank, p. 6’7 in TlzeMathematics of Dijjfiision,
Oxford Universitv Press. London. 1956.

P 34
12. H. J. de Nordwall, private communication. 4548, pp. 111 -18; see also Sect. 6, this report, %
13. C. E. Milstead, “Sorption Characteristics of the
Cesium-Graphite System at EIevatcd Temperatures and
Low Cesium Pressure,” Cnrhon (Oxford) 7, 199- 200
41969).
14. E. E. Compere and E. 6. Bohlmann, MSR
Program Serniamu. Progr. Rep. Feh. 28, 1990, ORNL-
especially Fig. 65.
15. A. Houtzeel and F. F. Dyer. R Study of Fissioiz
Products in the Molten Salt Reactor Expetfinent bey
Gamma Spectrometry. OKNGTM-3 15 1 (August 1972).

. . . . . . . . .
... . . . . . . :..*
.X$y//
.............
.. ... .
-&!.!,v
A detaiied synthesis of all the factors known to affect
fission product behavior in this reactor is not possible
within the available space. Many of the comments
which follow are based on a recent summary report.'
Operation of the MSRE provided an opportunity €or
studying the behavior of fission products in an
operating molten-salt reactor, and every effort was
made to maximize utilization of the facilities provided,
even though they were not originally designed for some
of the investigations which became of interest. Sig-
nificant difficulties stemmed f~om:
I.
2.
3.
4.
5.
The salt spray system in the pump bowl could not
be turned off. TIius the generation of bubbles and
salt mist was ever present; moreover, the effects
were not constant, since they were affected by salt
level, which varied continuously.
The design of the sampler system severely limited
the geometry of the sampling devices.
A mist shield enclosing the sampling point provided
a special environment.
Lubricating oil from the pump bearings entered the
pump bowl at a rate of I to 3 cm3/day.
There was continuously varying flow and blowback
of fuel salt between the pump bowl and an over-
flow tank.
In spite of these problems, useful information con-
cerning fission product fates in the MSRE was gAined.
135
12. SUMMARY AND OVERVIEW
12.1.1 Salt samples. The fission products Rb, Cs,
Sr, Ba, the lanthanides and Y, and Zr all form
stable fluorides which are soluble in fuel salt. These
fluorides would thereby be expected to be
found completely in the fuel salt except in those cases
where there is a noble-gas precursor of sufficiently long
half-life to be appreciabiy stripped to off-gas. Table
12.1 summarizes data from salt samples obtained during
the a u operation of the MSKE for fission products
with and without significant noble-gas precursors. As
expected, the isotopes with significant noble-gas pre-
cursors (89Sr and 13'Cs) show ratios to calculated
inventory appreciably lower than those without, which
generally scatter around or somewhat above 1 .O.
12.1.2 Ibeg~~ition. Stable fluorides showed little
tendancy to deposit on IIastelloy S or graphite.
Examinations of surveillance specimens exposed in the
core of the MSRE showed only 0.1 to 0.2% of the
isotopes without noble-gas precursors on graphite and
Hatselloy W. The bulk of the amount present stemmed
from fission recoils and was generally consistent with
the flus pattern.
However. the examination of profiles and deposit
intensities indicated that nuclides with noble-gas pre-
cursors were deposited within the graphite by the decay
of the noble gas that had diffused into the relatively
porous graphite. Clear indication was noted of a further
Table 12.1. Stable fluoride fission product activity as B fraction of cd~~lated
inventory in salt samples from 2 3 3 ~
operation
Without significant noble-gas precursor With noblegas precursor
Nuclide " ~ r 14'ce 144 Ce 14'Nd 89sP 137cs 91Y 1 4 0 ~ ~
Weighted yield, 70' 6.01 6.43 4.60 1.99 5.65 6.57 5.43 5.43
Half-life, days 65 33 284 11.1 52 30 yr. 58.8 12.8
Noble-gas precursor 89Kr ""9e 91Kr I4'xe
becursor half-life 3.2 rnin 3.9 rnin 9.8 sec 16 see
Activity in saltb
Runs 15-17 0.88-1.09 0.87-1.04 1.14-1.25 0.99-1.23 0.67-0.97 0.82-0.93 0.83-1.46 0.82-1.23
R M 1% ~
1 .05-1.09 0.95-0.99 1.16-1.36 0.82-1.30 0.84-0.89 0.86-0.99 1 .l6-1.55 1.10--1.20
Runs 19-20 0.95-1.02 0.89-1.04 1.17-1.28 1.10-1.34 0.70-0.95 0.81-0.98 1.13-1.42 1.02-1.20
aAIIocated fission yields: 93.2% 233U, 2.3% 235U, 4.5% 239Pu.
bAs fraction of calculated inventory. Range shown is 25-45 percentile of sample; thus half the sample values fall within this
range.

diffusion of the relatively volatile cesium isotopes, and
possibly also of Rb, Sr, and Ba; after production within
graphite.
12.1.3 Gas samples. Gas samples obtained from the
gas space in the pump bowl mist shield were consistent
with the above results for the salt-seeking isotopes with
and without noble-gas precursors. Table 12.1 shows the
percentages of these isotopes which were estimated to
be in the pump bowl stripping gas, based on the
amounts found in gas samples. Agreenient with ex-
pected amounts where there were strippable noble-gas
precursors is satisfactory considering the mist shieid,
contamination problems, and other experimental dif-
ficulties. Gamma spectrometer examination of the
off-gas line showed little activity due to salt-seeking
isotopes without noble-gas precursors. Examinations of
sections of the off-gas line also showed only small
amounts of these isotopes present.
12.2 Noble Metals
The so-called noble metals showed a tantalizingly
ubiquitous behavior in the MSRE, appearing as salt-
borne, gas-borne, and metal- arid graphite-penetrating
species. Studies of these species included isotopes of
Nb, hlo, Tc, Ru, Ag, Sb, and Te.
12.2.11 Salt-borne. The concentrations of five of the
noble-metal nuclides found in salt samples ranged from
fractions to tens of percent of inventory from sample to
sample. Also. the proportionate composition of these
isotopes remained relatively constant from sample to
sample in spite of the widely varying amounts found.
Silver-111 ~ which clearly wuuld be a metal in the MSRE
salt and has no volatile fluorides, followed the pattern
quite well and also was consistent in the gas samples.
This strongly supports the contention that we were
dealing with metal species.
These results suggest the following about the noble
metals in the MSRE.
The bulk of the noble metals remain accessible in
the circulating loop but with widely varying
amounts in circulation at any particular time.
In spite of this wide variation in the total amount
found in a particular sample, the proportional
composition is relatively constant, indicating that
the entire inventory is in substantial equilibrium
with the new material being produced.
The mobility of the pool of noble-metal material
suggests that deposits occur as an accumulation of
finely divided, well-mixed material rather than as a
“plate.”
No satisfactory correlation of noble-metal concentration
in the salt samples arid any operating parameter
could be found.
In order to obtain further understanding of this
particulate pool, the transport paths and lags of
noble-metal fission products in the hllSRE were examined
using all available data on the activity ratio of two
isotopes of the same element, 39.6-day ‘03Wu and
364-day “Ru. Data from graphite and metal surveillance
specimens exposed for various periods and
removed at various times, for niateriai taken from the
off-gas system, and for salt and gas samples arid other
materials exposed to pump bowl salt were compared
with appropriate inventory ratios and with values
calculated for indicated lags in a simple compartment
model. This model assumed that salt rapidly lust
ruthenium fission product formed in it, soale to
surfaces and most to a separate mobile “pool” of
noble-metal fission productt, presumably particulate or
colloidal and located to an appreciable extent in pump
bowl regions. Some of this “pool’9 material deposited
on surfi~ces and also appears to be the source of the
off-gas deposits. All materials sampled from or exposed
in the pump bowl appear to receive their ruthenium
activity jointly from the pool of retained material and
from more direct deposition as produced. Adequate
agreement of observed data with indications of the
model resulted when holdup periods of 45 to 90 days
were assumed.
12.22 Niobium. Niobium is the most susceptible of
the noble metals to oxidation should the U4’/U3+ ratio
be allowed to get too high. Apparently this happened at
the start of the U operation, as was indicated by a
relatively sharp rise in Cr2+ concentration; it was also
noted that 60 to about 100% of the calculated ”Nb
inventory was present in the salt samples. Additions of
a reducing agent (beryllium metal): which inhibited the
Cr2+ buildup, also resulted in the disappearance of the
‘ Xb from the salt. Subsequently the
Nb reappeared
in the salt several times for not always ascertainable
reasons and was caused to leave the salt by further
reducing additions. As the 23 I.J operations continued,
the percentage of ” Kb which reappeared decreased,
suggesting both reversible and irreversible sinks. The
”Nb data did not correlate ciosely with the Mo-Ru-Te
data discussed, nor was there any observable correlation
of its behavior with amounts found in gas sampless.
12-23 Gas-borne. Gas samples taken from the pump
bowl during the 235U operation indicated con-
centrations of noble metals that implied that substantial
percentages (30 to 100) of the noble metals being

..... ,:.;..~. -,.,
,. . . . . .
. ... ...... .
.i
W . d
produced in the MSRE fuel system were being carried
out in the 4 liters (STP)/niin helium purge gas. The data
obtained in the 23 3U operation with substantially
improved sanipling techniques indicated much lower
transfers to off-gas. In both cases it is assumed that the
noble-metal concentration in a gas sample obtained
inside the mist shield was the same as that in the gas
leaving the pump bowl proper. (The pump bowl was
designed to niinimize the, amount of mist in the
sampling area and also at the gas exit port.) It is our
belief that the * U period data are representative and
that the concentrations indicated by the gas samples
taken during U operation are anomalously high
because of contamination. This is supported by direct
examination of a section of the off-gas system after
completion of the 23 'U operation. The iarge amounts
of noble rnetais that would be expected on the basis of
the gas sample indications were not present. Appre-
ciable (1 0 to 1'7%) amorphous carbon was found in dust
samples recovered from the line, arid the amounts of
noble metals roughly correlated with the amounts of
carbon. This suggests the possibility of noble-metal
absorption during cracking of the oil.
In any event the gas transport of noble metals appears
to have been as constituents of particulates. Analysis of
Graphite
Metal
Surface
Fiow regime
Table 12.2. Relative deposition intensities fox rnobie metals
the deposition of flowing aerosols in conduits de-
veloped relationships between observable deposits and
flowing concentrations or fractions of production to
off-gas for diffusion and thermophoresis mechanisms.
The thermophoretic mechanism was indicated to be
dominant; the fraction of noble-metal production car-
ried into off-gas, based on this mechanism, was slight
(much below 1%).
12.3 Deposition in Graphite and Bastellsy N
The results from core suiveillance specimens and from
postoperation examination of components revealed
that differences in deposit intensity for noble metals
occurred as a result of flow conditions and that deposits
on metal were appreciably heavier than OK graphite,
paiticularly foi tellurium and its precursor antimony.
The find surveillance specimen array. exposed for the
last four months of MSRE operations, had graphite and
metal specimens matched as to configuration in varied
Row conditions. The relative deposition intensities (1 .O
if the entire inventory was spread evenly over all
surfaces) were as shown in Table 12.2.
The examination of some segments excised from
particular reactor components, including core metal and
graphite, pump bowl, and heat exchanger surfaces. one
Deposition intensity
9SNb 9gMo YPTc I03wu 106wLl 12sSb 129mTe L3ZTe
Surveillance specimens
Laminar 0.2 0.2 0.06 0.16
Turbulent 0.2 0.04 0.10
Laminar 0.3 0.5 0.1 0.3
Turbulent 0.3 1.3 0.1 0.3
Reactor components
Graphite
Core bar channel Turbulent
Bottom 0.54 0.07
Middle 1.09
Top 0.23
0.15
0.07
0.25 6.65 0.46'
1.06 1.90 0.9P
0.28 0.78 0.62'
Metal
Pump bowl Turbulent 0.26 0.73 0.27 0.38 2.85 0.89
Heat exchanger shell Turbulent 0.33 1.0 0.10 0.19 2.62 1.35'
Heat exchanger tube Turbulent 0.27 1.2 0.11 0.54 4.35 2.57'
Core
Rod thimble
Bottom
Middle
Turbulen t
Turbulent
1.42
1 .oo
1.23
0.73
1.54
0.58
0.50
0.42
3.27
1.35
1.65a
0.54'
~2127..
I e.
0.9
2.0

year after shutdown also revelaled appreciable accumu-
lation of these substances. The relative deposition
intensities at these locations are also shown in Table
12.2.
It is evident that net deposition generally was more
intense on metal than on graphite, and for metal was
more intense under more turbulent flow. Surface
roughmess had no apparent effect.
Extension to all the metal and graphite areas of the
system wauld require knowledge of the effects of flow
conditions in each region and the fraction of total area
represented by the region. (Overall, metal area was 26%
of the total and graphite 74%)
Flow effects have not been studied experimentally:
theoretical approaches based on atom mass transfer
through salt boundary layers, though a useful frame of
reference, do not in their usual form take into account
the formation, deposition, and release of fine particu-
late material such as that indicated to have been present
in the fuel system. Thus, much more must be learned
about the fates of noble metals in molten-salt reactors
before their effects on various operations can be
estimated reliably.
Although the noble metals are appreciably deposited
on graphite, they do not penetrate any more than the
salt-seeking fluorides without noble-gas precursors.
The more vigorous deposition of noble-metal nuclides
on Hastelloy N was indicated by postoperation
examination to include penetration into the metal to a
slight extent. Presumably this occurred along grain-
boundary cracks, a few mils deep, whch had developed
during extended operation, possibly because of the
deposited fission product tellurium.
12.4 Iodine
The salt samples indicated considerable I was not
present in the fuel, the middle quartiles of results
ranging from 45 to 71% of inventory with a median of
62%. The surveillance specimens and gas samples
accounted for less than 1% of the rest. The low
tellurium material balances suggest the remaining ’ I
was permanently removed from the fuel as I3’~e
(half-life, 25 min). Gamma spectrometer studies indicated
the *’ I formed in contact with the fuel returned
to it; thus the losses must have been to a region or
regions not in contact with fuel. This strongly suggests
off-gas, but the iodine and tellurium data from gas
samples and examinations of off-gas components do not
support such a loss path. Thus, of the order of
one-fourth to one-third of the iodine has not been
adequately accounted for.
It is recogniLed that, as shown in gas-coaled reactor
studies:-’ fission product iodine may be at paitial
pressures in off-gas helium that are too low for iodine
to be fmed by steel surfaces at temperatures above
about 400°C. However, various off-gas surfaces at or
downstream from the jumper line outlet were below
such temperatures and did not indicate appreciable
iodine deposition. Combined with low values in gas
samples, this indicates little iodine transport to off-gas.
12.5 Evaluation
The experience with the MSE showed that the noble
gases and stable fluorides behaved as expected based on
their chemistry . The noblemetal behavior and fates,
however, are still in part a matter of conjecture. Except
for niobium under unusually oxidizing conditions, it
seem clear that these elenients are present as metals
and that their ubiquitous properties stem from that fact
since metals are not wetted by, and have extremely low
solubilities in, molten-salt reactor fuels. Unfortunately
the MSRE observations probably were substantially
affected by the spray system, oil cracking products, and
flow to and from the overflow, a11 of which were
continuously changing, uncontrolled variables. The low
material balance on ’ ‘1 indicates appreciable undetermined
loss from the MSRE, probably as a noblenietal
precursor (Te, Sb).
Table 12.3 shows the estimated distribution of the
various fission products in a molten-salt reactor, based
on the MSRE studies. Unfortunately the wide variance
and poor materid balances for the data on noble metals
make it unrealistic to specify their fates more than
cpditati~ely . As a consequence, future reactor designs
must allow for encountering appreciable fractions of
the noble metals in all regions contacted by circulating
fuel. As indicated in the table, continuous chemical
processing and the processes finally chosen will substantially
affect the fates of many of the fission
products.
References
1. M. W. Rosenthal, P. K. Haubenreich, and R. B.
Briggs, The Developnzeizt Status of Molten Salt Breeder
Reactors, ORNL-4812 (August 15)72), especially chap.
5. “Fuel and Coolant Chemistry,” by W. R. Grimes, E.
6. Bohlmann, et al., pp. 95- 173.
2. E. Hoinkis. A Review of’ the Adsorption of Iodine
on Metal and Its Behavior in Loops, QRNL-IFM-2916
(May 1970).
3. C. E. Milstead, W. E. Bell, and J. H. Norman,
“Deposition of Iodine on Low Chromium-Alloy
Steels”, NucL Apple Techno!. 7, 361-66 (1969). .. ...
L

Table B2.3. Indicated distributiion of fission products in molten-salt reactors
- --- -..- ---
Fission product gmup Example isotopes
Distribution (%)
Hn salt To metal To graphite To off-gas Other
___-- -.
Stable salt seekers Zr-9.5, Ce-144, Nd-141 VB Negligible < 1 (fission recoils) Negligible RocessingQ
Stable salt seekers (noble gas precursors) Sr-89, Cs-I 37, Ba-140, Y-91 VariaI~le/T~~~ of gas Negligible Low Variable/T 1 ,2 of gas
Noble gases MI-89, Kr-91, Xe-135, Xe-137 Low/T1,2 Qf gas Negligible Low High/T,,, of gas
Noble metals Nb-9.5, MO-99, Ru-106, &-ill l-20 5-30 5-30 Negligible Processingh
Tellurium, antimony l-e-129, Te-121, Sb-125 l-20 20-90 5-30 Negligible Processingb
Iodine I-l 31, I-l 35 SO-IS

4. P. R. Rowland, W. E. Browning, and MI. Carlyle, 5. E. k. Conzpere, M. F. BS~UK~C, and H. J. deNord-
Behavior of Iodine Isotopes in a High Temperature Gas wall, Iodine Behavior in an HER, ORSL-TM-4744
Reactor Coolant Circuit, I). P. Report 736 (November
E 970).
(February 1975).
L4

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